ML20064N272

Proposed Tech Specs Addressing Changes Necessary to Support Fuel Cycles Designed W/Positive Moderator Temperature Coefficient

ML20064N272
Person / Time
Site:	Byron, Braidwood
Issue date:	03/23/1994
From:	COMMONWEALTH EDISON CO.
To:
Shared Package
ML19304B887	List: ML20064N257 ML20064N272 ML20064N321 ML20064N328 ML20064N341
References
NUDOCS 9403290239
Download: ML20064N272 (54)

Text

l t-u, e

ATTACHMENT 2 PROPOSED CHANGES TO APPENDIX A TECHNICAL SPECIFICATIONS FOR FACILITY OPERATING LICENSES NPF-37, NPF-66, NPF-72, and NPF-77 i

AFFECTED PAGE INDEX Byron Affected Pages: IV Braidwood Affected Pages: IV 2-2 2-2 B 2-1 B 2-1 2-5 2-5 2-7 2-7 3

2-8 2-8 2-10 2-10 3/4 1-4 3/4 1-4 3/41 Sa 3/41-Sa 3/4 1-11 3/4 1-11 3/4 1-12 3/4 1-12 3/4 2-8 3/4 2-8 3/4 5-1 3/4 5-1 3/4 5 11 3/4 5-11 3/4 9 -1 3/4 9 -1 B 3/4 1-2 B 3/4 1-2 83/41-3 83/41-3 B 3/4 2-4 83/42-4 83/45-4 B 3/4 5-4 83/46-3 B 3/4 6-3 B 3/4 9-1 B 3/4 9-1 j

i l

4 1

l

)

~

9403290239 940323 PDi:

ADOCK 05000454 P

PDR LJ

.t._

ATTACHMENT B (continued)

BYRON AFFECTED PAGES l

L1MITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.0 APPLICABILITY...............................................

3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL

~

Shutdown Margin - T,yg > 200*F...........................

3/4 1-1 Shutdown Margin - T,yg 3,200'F...........................

3/4 1-3 Moderator Temperature Coefficient........................

3/4 1-4

)

Minimum Temperature for Critica11ty......................

3/4 1-6 3/4.1.2 BORATION SYSTEMS Flow Path - Shutdown.....................................

3/4 1-7 I

Flow Paths - Operating...................................

3/4 1-8 Charging Pump - Shutdown.................................

3/4 1-9 r

i Charging Pumps - Operating...............................

3/4 1-10 Borated Water Source - Shutdown..........................

3/4 1-11 i

i Borated Water Sources - Operating........................

3/4 1-12 Baron Dilution Protection System........................

3/4 1-13a l

3/4.1.3 MOVABLE CONTROL ASSEMBLIES Group Height.............................................

3/4 1-14 TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN INOPERABLE FULL-LENGTH R00..............

3/4 1-16 Position Indication Systems - Operating..................

3/4 1-17 Position Indication System - Shutdown....................

3/4 1-18 Rod Drop Time............................................

3/4 1-19 Shutdown Rod Insertion Limit.............................

3/4 1-20 Control Rod Insertion Limits.............................

3/4 1-21 FIGURE 3.1-1 ROD BANK INSERTION LIMITS VERSUS THERMAL POWER FOUR LOOP 0PERATION...........................

3/4 1-22 F t c,u e E.5 l - O MdDCEATc2 Th(T%2A76eG ecEFFic1Ei1T

&c3us Th02 LEJEL 3/4 l-fa BYRDN - UNITS 1 & 2 IV Amendment No. 51

8pjag Qua Z.l-I with fy Z/-/ n

& na/ pay 670

............... -..._.. _._... -...._. j

....g._....-.-..3__..._.__

_.._.,..- ~. _ -

.__.___=.g

=.=t

._ __. :3 t=.

w 660. _ _. _

.._ _..._....- _. _.: ". _ _ _.. _.. =: = r

. _. _. _ _ _. _. _. =.

- -c;.__. __.

-- - -s

=. __ _;j 2 400 PS I A -'-

+.-;

650 ?__

N_ _. 5...._..__..~_~..l

- ' - ~

i

- _... _ 7

_. _. 3. \\.. :__ _

640

--2'25o Psia: "N w

%.s__. _ _.

N-

--,=

-r c:

3 N

i

_x

. \\-

L w

A g 630 g

N__

.s-j s

'--h' 5 ! _. ki_._.

W N

j

..-.i-..._

2000\\PS I A g

, _N_._ ___ m__, ____ i 7..w

- {s\\=E

~

y_.__.,...,_,

^R-f.___ ^-

- - ::1860 PS ta g

g x

....i**-

. "- 5 610.;_ g_=_. _ a..._,"==:--

t'.

5 3 =__..__. =-e

_._ s

\\ _. _ __ w_ -

' A._.

1___

. ; _..=-t _

k-

\\ _.

~

__.y q=

600

-~

~ _ _. _ _ _ _ _,

--3. m_,_

• =_

~

;;;-~,. g ; ;.;,;- ---j --. --. i

-.%-----4 L--

I-"**--_._..-

g "N.

• .\\

.3 O' f

-_"'_"=.;L=-~'-~':"._.~J-~~~=.;-~'_.-~~--'..__.._.:=._.==m=.==.==i.._.-_:-

=.

_ :===="

~=

-w

==:

= =!1 = 4:= =-_i= irr Is

=.=L r_ Za=_ s i", ="1... =:=.=.._,,1_h=. =.

...=-i==.=

=

.=

• ~YiE5I5.~iEE="if-?2 diira:I m m#8:EI."__=__ h =__-

=~-~-~C. $

!:':=U~"!E.

=:==.. _=_=_. _' i.*~ ~ ~.-~ P.ir:3E i =1.FfII.==.2;==. :t u. = E rl====C"

t' ; =.24__==)

.. 5._.=_3. 5_ _; =. _.. = : = :.=:_;.= - - _; ; =_- - _, - -.:g= :: =._:.-__. ; n - -

+ hr; =i :E j - :=.-i =

=lr=.. ; :== =p_- r.: rl gr.;.._ _:==isia.y 580 20 40 60 80 100 0

POWER (PERCENT)

FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION BYRON - UNITS 1 & 2 2-2

a 4

1 l

.\\

680 670 2471. psia.

.,.e.,..

..,.3 r- -..

3-c.-.-- r l

1 660

)

, i.

'22$$ pgj,31 r.-

.1r-t * - *.- -

1rr-N 650

^u Ov g

640 3

3 2 0 0 0. P. s. i.a.

m w

O C,

E 630

.w

..3

}C 1860' ps;La.....

O i.

CD

\\.

e 620

. x.

g

..'. J.. t.

. J.

1.. '.

4 A.

610

.1.'...'..

..'. J. L,

.J.t..'.

I 600 1

e i

s t

N N.

590 580 0

0.2 0.4 0.6 0.8 1

1.2 Power (Fraction of Nominal) l 1

1 l

i l

i i

Figure 2.1-1

{

Reactor Core Safety Limit - Four Loops 1

in Operation 1

i

t 2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission i

products to the reactor coolant.

Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the i

heat transfer coefficient is large and the cladding surface temperature is slightly aDove 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 3

related to DNB.

This relation has been developed to predict the DNB flux and

-l the location of DNB for axially uniform and nonuniform heat flux distributions.

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

Thm :50 J;;ign b;;i;,; a; fclicu;:

there me:t bc :t 'e?r+ ' 05 pa" eat

-p-ab:bi'i ty th:t the

-4 itu-DMSP Of the 'i-iting r:d dur#ng C;nditi;n !

nd'i!

i; gre:ter th:r er ;u:' to the DMBR ',4rit of th^ DMS corre,l:ticr being i

event:

.-_a r.u-

onn_,

n_.:_:

e.._

x____<,..

inens,.._

__a usww g wirw 4

e wswwivm av.

wyw.msswu s uwe

,~_11._m_u v,i__

_ _ _,s s wmv e j g wi sij ewwa mow w,iw m

mn

..,o m a

u. mm..n.e e i n. f.,_ m y y. 2, __ _ A 2,._ _ %

.m.r.

o nuy e

r.._i f_

TL.

_____,_t mm

.m 5Cf m,.

0" " '#-it ;; :;tabli;h;d b;;;d on th; catir; ;;plic;bi; exper ::nt:1 d;t:

t

][N5(trA cc,:b th:t ther: _

95 pe.r::nt pr:b:bility

't' 05 p:r ::t ::r'id:n;; th;t 2":

mm, m._,

,~e

%mm eum m, n numm unun u ow wnu m i c iaw wu unun s im w u.u avr u _

• u, i m, ~n.,

-_n m o

.L a

un, ~n _ o __

.3 1 3.s

.im

. _.. m. m o..,.

In ::: ting thi; de:igr b::i;, un::rt:inti::

4-n

? :rt cpe":ti g p="'~a+e":,

inn-les, 2na +horms1 73r3 meter, 2nd f;;] f gri;;;j;r p;r:reter cre rentidered

-Ot:tistically ;uct th;t thcr; i; at 1;;;t ; ^ confid;n;; that th; mini;;; 0"""

for th: '4-iting red i: greater th;n ;r equ:1 te th: DNBP 14-4t The u re--

1;intic: ;m th; abo;; plant par;;;t;r; cr; ;;;d to d;tcrr,in; th: ;':rt DME" uncertainty.

'hi; DNSD un ;rtainty, c;abH,;d with th; ;;rr lation DMC" limit, c;tabli;ht; ; d::igr DMSR ;;1u; which mu;t b; ;;t in plant ;;f:ty an:ly:i:

u:ing value: Of i nput par : ter; ithout un rtainti::

I. Ls1 a n d I 20I The design DNBR values are, 1.21 nd 1.22 for a typical, cell and a thimble ce11, respec, tive1y,,. _ _.

n-,

mm m,

,,,-m_

m

_2__,

mo mue......

m yr..

-. ~........

t3j;t1: ::1

+we vouTacy c

<ns, In addition, margin has been maintained g:5 in both designs by meeting safety analysis DNBR limits of-tr+grtor a typical cell and,..,,,,

,,m thimble cell zm.

e, m m,.,

. -. y r..,.

ne, e,

._2

, ee,._...:

m mo mm..., -~....

l' nd trimbl: ::11, r;;p;;ti.;13 f;; th; 'la" TACE : fu:1 in performing i

safety analyses.

The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum design DNBR is no less than the design DNBR value, or the average enthalpy at th.e vessel exit is less than the enthalpy of saturated liquid.

BYRON - UNITS 1 & 2 B 2-1 AMENDMENT NO. 36

Insert A The DNBR thermal design criterion.is that the probability that DNB will not occur' on the most limiting rod is at least 95% (at a 4

95% confidence level)'for any Condition I or II event.

In meeting this design basis, uncertainties in plant operating parameters, nuclear and thermal ~ parameters, and fuel fabrication parameters are considered.

As described in the UFSAR, the effects of these uncertainties have been statistically combined with the correlation uncertainty.

Design limit DNBR values have been determined that satisfy the DNB design criterion.

f e

t TABLE 2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS h

FUNCTIONAL UNIT TRIP SETPOINTS All0WABLE YALUE

=

3

12. Reactor Coolant Flow-Low 190% of loop mint-189.3% of loop alpi-mum measured flow mum measured flow

[

13. Steam Generator Water level Low-low a.

Unit 1 233.0% of narrow 231.0% of narrow range instrument range instrument b.

Unit 2 span span 236.3% of narrow 234.8% of narrow range instrument range instrument span span w

14. Undervoltage - Reactor 25268 volts -

24920 volts -

/n Coolant Pumps each bus each bus

15. Underfrequency - Reactor 257.0 Hz 256.08 Hz Coolant Pumps

16. Turbine Trip a.

Emergency Trip Header 11000 psig 1815 psig Pressure b.

Turbine Throttle Valve 21% open 11% open Closure k

17. Safety Injection Input N.A.

H.A.

g from ESF

18. Reactor Coolant Pump N.A.

N.A.

Breaker Position Trip

• Minimum measured flow - 07,'00 gpm 92,85e

k 6.

I c

yc fo 5k h

k

))

e I

s yg l

A s

(

8 0

f

=

g

=

g v

c p

3 y

a r

3 T

l

)

1 T

u J,

i

'P r

T r

o A

T o

f k

A f

l a

r r

P o

r r

o a2

(

f o

o t

a f

t a

K r

21 a

s o

r s

n

+

t o

e p

n e

c a

t

]

n s

a p

m 4

o n

s m

o Mu

'T i

e n

o c

t p

e c

a m

p g

U e t

o m

g a

)

n c

o a

l 5

e c

l o

e 4

m T

g R

d

)

r l

a W

. d a

'd u

A a

g E

l A

a e

d I

+

t d

l O

e i

g e

e s

e d

P l

u 1

n r

a T

e e

c.

l n

(

I u

e A

h L

e h

T b

i S

s l

t A

h t

j t

N T

d a

d M

. t d

a p

n O

(

l e

n e

n R

n e

lc p,

o I

o m

i r

i E

y i

r C

T

))

f u

b u

l pA l

(

A 5, 5 i

n d

s d

T in o

e a

e d

F s

T 9

a z

e z

D

. d e

a p

1 O

1 e,

z e

N H

r i

m 2

+ +

o l

l T

so l

e, m h

i E

t n i

1 E

D t

i n

i A

ri i

r n,

l 2

L 1O T

a t

o t

R et t

' u o

4 B

R s

u u

na u

t

(

1i E

A n

r t

es a

r n

L T

y e

s o

s a

gn s

r o

i B

K b

p t

t t

e t

e t

t L

A m

n a

n T

np n

p a

T T

o a

s a

A i

om a

m s

f io t

e n

n o

A c

t n

t K'

d g

ns p

n e

l s,

e s

d f

t c s,

t e

U g

c ns p

(

e a

o m

o t

/

nc o

'.e m

e r

l c3 o

c a

}

5 ui c4 g

o

'o l

o u

c c

4 6

f m

,a c

b s

d m=

e e

i 6

2 a

n=

d e

r a

a g

m d

1 a

0 en e

g e

e i 2 a

i n

h y i 3 v

a 5

H L

T1 L

T I

1 0

Td T r A

L i i r

(

(

}

b l

p S

=

=

=

=

=

=

=

==

=

=

=

a p

a I

c/

i T 1 3

l 3,

pf A

2 s

1 1

p o

i E

(

1 d

R 1

A T"

U

))

T a

3 g

4 T S5 A

I t

t g

1 A

K K

I g

1 T

1 4

r A

i 2

4h R

t t

w E

P

• 4 T

M W

e)

E 1 1

e S

z I

((

r R

e Z3 o.

E T

h V

A W

o D

s 1

E T

O N

q y, E d w e-o 7

t

g TABLE 2.2-1 (Continued) we TABLE NOTATIONS (Continued) g NOTE 1:

(Continued) b r,

- Time constant uttilzed in the measured T,, lag compensator, t, - O s, T'

s 588.4*F (Nominal T.,at RATED THERNAL POWER),

o.

x to o.ooi81 K

-(D.00134[*

3 P

Pressurtzer pressure, psig, P'

2235 psig (Nominal RCS operating pressure),

Laplace transform operator, s,

S and f,(AI) is a function of the indicated difference between top and bottom detectors of the power-range neutron ton chambers; with gains to be selected based on measured instrument responseduringplantSTARTgPtestssu_chthat:.,,,,y to

- Z4 %

E.

(1) for q - q, betweed-32%)u 1and G13%ff (&I) - 0, where q and q are percent RATED THERMAL l

POWER,in the top and bottom halves,of the core respec,tively,, and q, + q,, is total THERMAL POWER in percent of RATED THERMAL POWER;,,g.g (11) for each percent that the magnitude of q - q exceeds 413%fthe AT Trip Setpoint shall be automatically reduced by 0.74%foi its,valu,e at RATED THERMAL POWER.

b4.11%"

- 2 4 */o

• be automatically reduced by(gnitude of q, - 4 exceedsk32%).

(111) for each percent that the ma the AT Trip Setpoint shall C.67%)M lts value at RATED THERMAL POWER l

5.35%"

g NOTE 2: *(The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than m

g i.tc% 3.71%)"o'r dT spaa.

l b

5 Aeplicabl< 4e uni 4 L. 4plicabl< 4c, ual Z.afhr eyela 5.

10 tb4 Applicable 40 un(t i. Applicab l<. 4c, ud4Z 4;I expe4.'ew of e9eI, 5.

l

.4

I Q

TABLE 2.2-1 (Continued) l TABLE NOTATIONS (Continued) e NOTE 3:

(Continued)

' O 00 a

[

K.

= (0.00170/'F)for T > 1" and K = 0 for T i T",

es *

[

T

=

As defined in Note 1, 1"

Indicated T,yg at RATED THERMAL POWER (Calibration temperature for ai

=

instrumentation, 3 588.4*F),

S

=

As defined in Hote 1, and f (al) 0 for all al.

=

ty NOTE 4:

The ch 4

02.31%)ggnel's maximum Trip setpoint shall not exceed its computed Trip Setpoint by more than So8%

g or AT span.

W - k pp se d.ble do dos'I 1.

hlsectble h Linil. [ g[4 str l

d I(

d3

• • _ a t Appi,c at 4e u e t. 1.

Ql. cable % untt z. uAt omi<6 t

a of ge t< 5-

?

e

REACTIVTTY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION

)

3.1.1.3 The moderator temperaturi coefficient (MTC) shall be:

a.2[less positive than 0 ak/k/ F for the all rods withdrawn, hot zero l

THERMAL POWER condition, g b.

Less negative than -4.1 x 10 4 ok/k/ F for the all rods withdrawn, end of cycle life (E0L), RATED THERMAL POWER condition.

i APPLICABILITY:

Specification 3.1.1,3a. - MODES 1 and 2* only#,

Specification 3.1.1.3b. - MODES 1, 2, and 3 only#.

ACTION:

With the MTC more positive than the limit of Specification 3.1.1.3a.

a.

above, operation in MODES 1 and 2 may proceed provided:

1.

Control rod withdrawal limits are established and maintained sufficient to restore the MTC to less positive thaq ok/k/F)u within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in HOT STANDBY within the nex 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

These withdrawal limits shall be in addition to the insertion limits of Specification 3.1.3.6;

.%< lieis of u

2.

The control rods are maintained within the withdrawal limits established above until a subsequent calculation verifies that I

the MTC has been restored to within its limit for the all rods withdrawn condition; and i

3.

A Special Report is prepared and submitted to the Commission pursuant to Specification 6.9.2 within 10 days, describing the k

value of the measured MTC, the interim control rod withdrawal limits, and the predicted average core burnup necessary for

{

restoring the positive MTC to within its limit for the all rods withdrawn condition.

4.

The provisions of Specification 3.0.4 are not applicable.

l b.

With the MTC more negative than the limit of Specification 3.1.1.3b.

above, be in HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, a.l" dainbuJ ehi 4kn IMis specifi<J 'm K ure 5. l - O,

3

• With Keff greater than or equal to 1.

1. See Special Test Exceptions Specification 3.10.3.

• < App!;eable 4e Uni 41. Applicable lo ud t Z. All4 < c') e 18 d-

1. Lle4 Appliu bt, 4e u a t 1.

A pllem61,40 OMI z u 4; \\ ocup) un(4 2 J[e, ege [, 5.

i

REACTIVXTY CONTROL SYSTEMS BORATED WATER SOURCE - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.5 As a minimum, one of the following borated water sources shall be OPERABLE:

A Boric Acid Storage System with:

a.

1)

A minimum contained borated water level of 7.0%,

2)

A minimum boron concentration of 7000 ppe, and 3)

A minimum solution temperature of 65'F.

b.

The refueling water storage tank (RWST) with:

1)

O af. A minimum contained borated water level of 9.0%,

A bem ene,a4voE bekrA 45o0M de ppw,

-@bfA minimum boron concentration of 2000 ppa, and 3)

A minimum solution temperature of 35'F.

APPLICABILITY: MODES 5 and 6.

ACTION:

With no borated water source OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

SURVEILLANCE REOUIREMENTS 4.1.2.5 The above required borated water source shall be demonstrated OPERABLE:

At least once per 7 days by:

a.

1)

Verifying the boron concentration of the water, 2)

Verifying the contained borated water' level' and 3)

Verifying the boric acid storage tank solution temperature when it is the source of borated water.

b.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying the RWST temperature when it is the source of borated water and the outside air temperature is less than 35'F.

n V-able k Ual i.. hep l; cable.Se, uu L Z. a k ey e le S.

BYRON - UNITS 1 & 2 3/4 1-11 k4 dod lLCCibk(

d d$

b.

p(ICCh.

C N

glc4b ofc9clc 5.

WA

. ~..

REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.6 As a minimum, the following borated water source (s) shall be OPERABLE as required by Specification 3.1.2.2 for MODES 1, 2 and 3 and one of the following borated water sources shall be OPERABLE as required by Specification y

3.1.2.1 for MODE 4:

)

f a.

A Boric Acid Storage System with:

)

1)

A minimum contained borated water level of 40%,

2)

A minimum baron concentration of 7000 ppa, and 3)

A minimum solution temperature of 65'F.

b.

The refueling water storage' tank (RWST) with:

1) / minimum contained borated water level of 89%,

-E-) bf'l minimum boron concentration of 2000 ppa,be onem z) a) 1 3)

A minimum solution temperature of 35'F, and

' l 4)

A maximum solution temperature of 100*F.

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

ACTION:

With the Boric Acid Storage System' inoperable and being used as one of '

a.

the above required borated water sources in MODE 1, 2, or 3, restore the system to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at'least HOT..

STANDBY'within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and borated to a SHlITDOWN MARGIN equivalent to at least 1% Ak/k at 200*F; restore the Boric Acid Storage System to OPERABLE status within the next 7 days or be in

- l COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b.

With the RWST inoperable in MODE 1, 2, or 3, restore the tank to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least HOT STANDBY within-the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

With no borated water source OPERABLE in MODE 4, restore one borated c.

water source to OPERABLE status within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or be in COLD SHlITDOWN - I within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

j l

• Appheable de udt. t. Appli ca.ble 4e u d4 2, A g of, 5

" Lt.4 e.pplimbt, 4e us11. $ppheable 4e udt1 A\\ - coupleG BYRON - UNITS 1 & 2 of OJ' e U I-3/4 1 12 m

POWER DISTRIBUTION LIMITS

~

3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR LIMITING CONDITION FOR OPERATION 3.2.3 Indicated Reactor Coolant System (RCS) total flow rate and F"# shall be maintained as follows for four loop operation.

378400SP*

RCS Total Flowrate > 290,,^00==, and a.

Fh51.55[1.0+0.3(1.0-P)]for0FAfuel b.

Fh11.65[1.0+0.3(1.0-P))forVANTAGE5 fuel where:

MeasuredvaluesofFhareobtainedbyusingthemovableincore detectors. An appropriate uncertainty of 4% (nominal) or greater shall then be applied to the measured value of Fh before it is compared to the requirements, and, THERMAL POWER p _ RATED THERMAL POWER APPLICABILITY:

MODE 1.

ACTION:

WithRCStotalflowrateorFhoutsidetheregionofacceptableoperation:

a.

Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> either:

N 1.

Restore RCS total flow rate and F to within the above limits, M

or 2.

Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux-High Trip Setpoint to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

BYRON - UNITS 1 & 2 3/4 2-8 AMENDMENT NO. 36

e 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3 /4. 5.1 ACCUMULATORS LIMITING CONDITION FOR OPERATION 3.5.1 Each Reactor Coolant System accumulator shall be OPERABLE with:

The isolation valve open and power removed, a.

b.

$ contained borated water level of between 31% and 63%,

W

~.1) A b:nu.ce*** bah t>chocen 22co and loco ppw.

c A 2ffboron concentration of between 1900 and 2100 ppm, and d.

A nitrogen cover-pressure of between 602 and 647 psig.

APPLICABILITY: MODES 1, 2, and 3*.

ACTION:

With one accumulator inoperable, except as a result of a closed a.

isolation valve, restore the inoperable accumulator to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b.

With one accumulator inoperable due to the isolation valve being closed, either immediately open the isolation valve or be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.5.1.1 Each accumulator shall be demonstrated OPERABLE:

At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by:

a.

1)

Verifying the contained borated water level and nitrogen l

cover pressure in the tanks, and 2)

Verifying that each accumulator isolation valve is open.

^ Pressurizer pressure above 1000 psig, kppl odLbic k lini4.i.,

kppls'ca ble k UA b 2 B b agolt 6.

• • uo4 Appi,eAl< -lo uM41. Applica.bl< 4e uti4 2 _k t ompiclis of %c tL N-BYRON - UNITS 1 & 2 3/4 5-1 AMENDMENT NO. 21

_1f EMERGENCY' CORE COOLING SYSTDts 3/4.5.5 REFUELING WATER STORAGE TANK LIMITINGCONDITIONFOR'OPikATION 3.5.5 The refueling water storage tank (RWST) and the heat traced portion of the RWST vent path shall be OPERABLE with:

b.1) A minimum contained borated water level of SSE, a.

A w - **

• W h>ua 25e w 25eo epm, 4 z M sinimum boron concentration of.2000 ppe, c.

A minimum water temperature of 35'F, and-d.

A anximum water temperature of 100*F.

APPLICABILITY: MODES 1, 2, 3, and14.

ACTION:

With the RWST inoperable, restore the tank to OPERABLE status within 1 hour-or' be in at least NOT STANDtY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the-following-30 hours.

SURVEILLANCE REQUIREMENTS 4.5.5 The RWST shall be demonstrated OPERABLE:

a.

At least once per 7 days by:

1)

Verifying the contained borated water level in the. tank, and 2)

. Verifying the boron concentration of the water, b.

At least ence per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying the RWST temperature when the outside air temperature is'either less than 35'F or greater than 100*F, and-At.least' once per 24 heurs by verifying' the RWST vent path -

c.

temperature to be greater than or e air temperature is less than 35'F. qual to 35'F when the outside

• • _ Alo4 Applica.ble -lo ud i-1 4pplicabt< 4o thik Z J I coup d i

ofc3cic5.

SYRON --UNITS 1.4 2 3/4 5-11 AMENDMENT NO. 38 :

3/4.9 REFUEL 7NG OPERATIONS 3/4.9.1 BORON CONCENTRATION LIMITING CONDITION FOR OPERATION

~,

3.9.1 The boron concentration of all filled portions of the Reactor Coolant System and the refueling canal shall be maintained uniform and sufficient to ensure that the more restrictive of the following reactivi.ty conditions is

~

met:

A K,ff of 0.95 or less, or a.

b.

A boron concentration of greater than or equal to-0000 ppm.

APPLICABILITY:

MODE 6*.

ACTION:

With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm boron or its equiv-alent until K,ff -is reduced to less than or equal to 0.95 or the boron concerytration is restored to greater than or equal to G400-ppm, whichever is the more restrictive.

g SURVEILLANCE REQUIREMENTS s

4. 9.1.1 The more restrictive of the above two reactivity conditions shall be determined prior to:

a.

Removing or unbolting the reactor vessel head, and b.

Withdrawal of any full-length control rod in excess of 57 steps (approximately 3 feet) from its fully inserted position within the reactor vessel.

4.9.1.2 The boron concentration of the Reactor Coolant System and the refueling-canal shall be determined by chemical analysis at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

4.9.1,3 Valves'.CV111B,'CV8428,'CV8441,;CV8435, and CVB439 shall be verified closed and secured in position by mechanical stops or by removal of air or electrical power at least once per 31 days.

^The reactor shall be maintained in MODE 6 whenever fuel is in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.

BYRON - UNITS 1 & 2 3/4 9-1

~REACTIV!TY CONTROL SYSTEMS BASES MODERATOR TEMPERATURE COEFFICIENT (Continued)

The most negative MTC value equivalent to the most positive moderator density coef ficient (MDC), was obtained by incrementally correcting the MDC useo in tne FSAR analyses to nominal operating conditions. These corrections involved subtracting the incremental change in the MDC associated with a core condition of all roas inserted (most positive MDC) to an all rods withdrawn condition and, a conversion for the rate of change of moderator density with temoerature at RATED THERMAL POWER conditions.

This value of the MDC was then-transformed into the limiting MTC value -4.1 x 10 4 ok/k/*F.

The MTC value of -3.2 x 10 4 ak/k/*F represents a conservative value (with corrections for burnno and soluble boron) at a core condition of 300 ppm equilibrium boron concentration and is obtained by making these corrections to the limiting MTC value of -4.1 x 10

• ak/k/*F.

The Surveillance Requirements for measurement of the MTC at the beginning and near the end of the fuel cycle are adequate to confirm that the MTC can be maintained within its limits.

The BOL MTC measurement, combined with the

~

predicted MTC throughout core life, will be used to impose administrative limits on rod withdrawal, as required juring core life to ensure that MTC will always be less positive than principally to the reduction;O AK/K/*FT This coefficient changes slowly due

'in RCS boron concentration associated with fuel burnup.

L + 7.o v. e #4t'/uq- (~ 4.nl roJs wit %am L powv_ ltocl3

% 7e%EMDWtPWM.bEU with A limar re.g l*

3 /4.1.1. 4 MINIMUM TEMPERATURE FOR CRITICALITY oarkmd ke% EMTWE%

This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 550*F.

This limitation is required to ensure: (1) the moderator temperature coefficient 15.

within its analyzed temperature range, (2) the trip instrumentation is within its normal operating range, (3) the pressurizer is capable of being in an OPERABLE status with a steam bubble, (4) the reactor vessel is above its minimum RTNDT temperature, and (5) the plant is above the cooldown steam dump permissive, P-12.

3/4.1.2 BORATION SYSTEMS The Baron Injection System ensures that. negative reactivity control is available during each MODE of facility operation.

The components required to perform this function include: (1) borated water sources, (2) charging pumps (3) separate flow paths, (4) boric acid transfer pumps, and (5) an emergency, power supply from OPERABLE diesel generators.

t With the RCS average temperature above 350*F, a minimum of two baron injection flow paths are required to ensure single functional capability in the event an assumed failure renders one of the flow paths inoperable.

The boration capability of either flow path is sufficient to provide a SHUTDOWN MARGIN from expected operating conditions of 1.3% ak/k after xenon decay 'and cooldown to 200*F.

The maximum expected boration capability requirement is occur; et ECL 'r - fuP p;;r equilibrian unca condition; :nd r quire; t 5,4ea

(.15,780$allons of 7000 ppm borated water from the boric' acid storage tanks or.

gog (J0,450) gallons ofg000-ppmfborated water from the refueling water storage tank.

L I5m ppm BYRON - UNITS 1 & 2 B 3/4 1-2 AMENDMENT NO.36 r bl. 4pplioWe do Udt i, Appl ca.ble de lwd 2. uM coy

• [ Cfl

REACTIVITY CONTROL SYSTEMS BASES BORATION SYSTEMS (Continued) g4487 A Boric Acid Storage Sysg(em leve) of 40% ensures that there is a volume of greater than or equal to 15,780) gallons available. A RWST level of 89% ensures that there is a volume of greater than or equal to 395,000 gallons available.

With the RCS temperature below 350'F, one Boron Injection System is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single Boron Ir.jection System becomes inoperable.

The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps except the required OPERABLE pump to be inoperable below 330'F provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV or an RHR Suction,valv,e,q g The baron capability required below 200*F is sufficient to provide a SHUTDOWN MARGIN of 1% Ak/k after xenon decay gnd cooldown from 200*F to N

146*F. ThTs condition requires either(2,652) gallons of 7000 ppm bora d water from the boric acid storage tanks or41,840) gallons ofd2000 ppm) orated _zscoppm ZM water from the refueling water storage tank (RWST).

A Boric Acid StorJa e System level of 7% ensures there is a volume of greater than or equal to.Q652Ra11onq4o available.

An RWST level of 9% ensures there is a volume of greater than or equal to 38,740 gallons available.

The contained water volume limits include allowance for water not available because of discharge line location and other physical characteristics.

r So The limits on contained water 4 volume and baron concentration of the RWST also ensure a pH value of between-e-& and 11.0 for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.

The OPERABILITY of one Boron Injection System during REFUELING ensures that this system is available for reactivity control while in H00E 6.

The OPERABILITY of the automatic Boron Dilution Protection System ensures adequate capability for negative reactivity insertion to prevent a transient caused by the uncontrolled dilution of the RCS in MODES 3,4, and 5.

The func-tioning of the system precludes the necessity of operator action to prevent fur-ther dilution by terminating flow to the charging pump (s) from possible unborated water sources and initiating flow from the RWST. The most restrictive condition occurs shortly after beginning of life when the critical boron concentration is highest, and a 205 gpm dilution flowrate provides the maximum positive reactivity addition rate. One reactor coolant pump in operation with all reactor coolant loop stop isolation valves open reduces the reactivity addition rate by mixing the dilution through all four reactor coolant loops.

A minimum count rate of ten counts per second minimizes the impact of the uncertainties associated with the source range nuclear instrumentation.

In the analysis of this accident, a minimum SHUTDOWN MARGIN of 1.3 Ak/k is required to control the reactivity tran-sient. Actions taken by the micrtprocessor if the neutron count rate is doubled will prevent return to criticality in these MODES.

1 BYRON - UNITS 1 & 2 B 3/4 1-3 AMENDMENT NO. 51 1

dd Applicable b Lidi 1.. blicable. Jo 044 2 udl ampida c[ cycle 6 E

POWER DISTRfBUTION LIMfTS

}

BASES-HEAT FLUX HOT CHANNEL FACTOR. and RCS FLOW RATE AND NUCLEAR ENTHALPY RI

-HOT CHANNEL FACTOR (Continueo)

The co'ntrol rod insertion limits of Specification 3.1.3.6 are c.

maintained, and d.

The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.

H F

will be maintained within its limits provided the ' Conditions a. thr h

ON

d. above are maintained.N Th'e combination of the RCS flow requirement'( d % f.... _. _

SP and the requirement on F will be met.

3g guarantee that the DNBR used in the safety analysis 1.5o Margin between the safety analysis limit DNBRs (1.1^

c4 1.07 fe, tL Ort.

Y typic:' rdt"-t!: nll:, r :; nti;;l; :nd 1.07 :nd 1.05 for-the 1.25 -

" N E 5 typical and thimble cells) and the design limit DNBRs (1. :

1. :

f;c n; CP' f _:1 typin! =d tP:M: n il;, x

1. 22 :nd 1. 2 for the

'/l,NTl,0 ; ful typical and thimble cellsg..;,.. ni..lj) is maintained.

'i A fraction of this margin is utilized to accommodate the t x;iti: r :

ON;n gnait3 ',;xix cf 12.5*') =d th: appropriate fuel rod bow DNBR penalty (less than 1.5% per WCAP-8691, Revision 1). The rest of the margin between design and safety analysis DNBR limits can be'used for plant design flexibility.

p, g 92,85o The RCS flow requirement is based on the loop minimum measured flow rate g

of 0.,000 gpm which is'used in the h;pr nd Thermal Design Procedure.d:= r!M d

" P ' ' 1 rd 15 0.2.

and is used to calibrate the RCS flow rate indicators.A precision heat balance Potential fouling of the feedwater venturi, which might not be detected, could bias the.results from the precision heat balance in a non-conservative manner.

Therefore, a' penalty of 0.1% is assessed for potential feedwater venturi fouling.. A maximum measurement uncertainty of M has been included in the loop minimum measured flow rate to account for potential undetected feedwater venturi fouling and the use of the RCS flow indicators for flow rate verification.

Any fouling which might bias the RCS flow rate measurement greater than 0.1% can be detected by monitoring and trending various plant performance parameters.

If detected action shall be taken, before performing subsequent precision heat balance measu,rements. i.e.,

either the effect of fouling shall be quantified and compensated for in the RCS flow rate measurement, or the venturi shall be cleaned to eliminate the fouling.

Surveillance Requirement 4.2.3.4 provides~ adequate monitoring to detect possible flow reductions due to'any. rapid cora crud buildup.

Surveillance Requirement 4.2.3.5 specifies that the measurement instrumen -

tation shall be calibrated within seven days prior to the performance of the calorimetric flow measurement.

This requirement is due to the fact that the.

drift effects'of this instrumentation are not included in the flow measurem uncertainty analysis.

This requirement does not apply for the instrumentation whose drift effects have been included in'the uncertainty analysis.

4 a

BYRON - UNITS 1 & 2 B 3/4 2-4 AMENDMENT NO. 36

~.

EMERGENCY CORE COOLING SYSTEMS BASES 3/4.5.5 REFUELING WATER STORAGE TANK The OPERABILITY of the refueling water storage tank (RWST) as part of the.

ECCS ensures that a sufficient supply of borated water is available for injection by the ECCS in the event of a LOCA. The limits on RWST minimum volume and boron concentration ensure that: (1) sufficient water is available within containment to permit recirculation cooling flow to the core, and (2) the reactor will remain suberitical in the cold condition following mixing of the RWST and the RCS water:

volumes with all control rods inserted except for the most reactive control assembly. These assumptions era consistent with the LDCA analyses.

The contained water volume limit includes an allowance for water not, usable because of tank discharge line location or other physical characteristics.

A minimum contained borated water level of 895 ansures a volume of greater than-or equal to 395,000 gallons.

The limits on contained water ' volume and boron concantration of the RWST also ensure a pH value of betweenHh4 snd 11.0 for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution of.

iodina and minimizes the effect of chloride and caustic stress corrosion on sechanical systems and components.

I

')

l I

BYRON - UNITS 1 & 2 8 3/4 5-4 AMENDMENT No. 38 re

CONTAINMENT SYSTEMS BASES CONTAINMENT PURGE VENTILATION SYSTEM (Continued) be exceeded in the event of an accident during containment purging operation.

Operation with one line open will be limited to 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> during a. calendar year.

Leakage integrity tests with a maximum allowable. leakage' rate for containment purge supply and exhaust. supply valves w~ill provide early indica-tion of resilient material seal degrad& tion' and will' allow impportunity for" m o repair before gross leakage failures could develop. 'The 0.60~ L ' leakage lintt i of Specification 3.6.1.2.b. shall not be exceeded when the-leak $ge rates determined by the leakage integrity tests of these valves are added to the previously determined total for all valves and penetrations subject to Type B and C tests.

3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS ta;

.m m.. :.

, 1.,..,, s.

3/4.6.2.1 CONTAINMENT SPRAY SYSTEM'

~'... "...

The OPERABILITY of the Containment S'p'r'ay System; ensures that containment.

depressurization and cooling capability will' be available in 'the'evint'of.a' ~ -

LOCA or steam line break. The pressure reduction-and resultant.lowe.r. containment leakage rate are consistent with the assumptions used in the safety, analyses.

The Containment Spray System.and the Containment. Coc1Jngafys redundant to each other in providing-. post-accident: cooling of the@co,ntain, ment a3 p' 7,.

atmosphere.

However, the Containment Spray System also provides a mechanism-for removing iodine from the containment atmosphere and therefore the time requirements for restoring an inoperable Spray System to OPERABLE status have been maintained consistent with that assigned other inoperable ESF equipment.

3/4.6.2.2 SPRAY ADDITIVE SYSTEM g

The OPERABILITY of the Spray Additive System ensures that. sufficient NaOH' is added to the containment spray in the event of a LOCA.

The limits on NaOH volume and concentration ensure a pH value of between nd 11.0 for the solution recirculated'within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.

The contained solution volume-limit includes an allowance for solution not usable be'cause of tank discharge line location or other physical characteristics.

These assumptions are consistent with the iodine removal efficiency assumed in the safety analyses.

A spray additive tank level of between 78.6% and 90.3% ensures a volume of greater than or equal to 4000 gallons but less than or equal to 4540 gallons.

BYRON - UNITS 1 & 2 B 3/4 6-3

9 l

3/4.9 RuFUELING OPERATIONS l

BASES 3/4.9.I BORON CONCENTRATION The limitations on reactivity conditions during REFUELING ensure that:

(1) the reactor will remain suberitical during CORE ALTERATIONS, and (2) a uniform boron concentration is maintained for reactivity control in the water volume having direct access to the reactor vessel. The limitation on Keff of no greater than 0.95 is sufficient to prevent reactor criticality during refueling operations and includes a 1% Ak/k conservative allowance for 2%5xo uncertainties.

Similarly, the boron concentration value of-9996' ppm or greater includes a conservative uncertainty allowance of 50 ppa. These limitations are consistent with the initial conditions assumed for the boron dilution incident in the safety analyses. The locking closed of the required valves during refueling operations precludes the possibility of uncontrolled boron -

dilution of the filled portions of the RCS. This action prevents flow to the RCS of unborated water by closing flow paths from sources of ur. borated water.

3/4.9.2 INSTRUMENTATION The OPERABILITY of the Source Range Neutron Flux Monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core.

3/4.9.3 DECAY TIME The minimum requirement for reactor suberiticality prior to movement of irradiated fuel assemblies in the reactor vessel ensures that sufficient time has elapsed to allow the radioactive decay of the short-lived fission products.

This decay time is consistent with the assumptions used in the safety analyses.

3/4.9.4 CONTAINMENT BUILDING PENETRATIONS The requirements on containment building penetration closure and OPERABILITY ensure that a release of radioactive material within containment wi.ll be restricted from leakage to the environment. The OPERABILITY and closure restrictions are sufficient to restrict radioactive material release from a fuel element rupture based upon the lack of containment pressurization potential while in the REFUELING MODE.

The Byron Station is designed such that the containment opens into the fuel building through the personnel hatch or equipment hate'h.

In the event of a fuel drop accident in the containment, any gaseous radioactivity escaping from the containment building will be filtered through the Fuel Handling Building Exhaust Ventilation System.

3/4.9.5 COMMUNICATIONS The requirement for communications capability ensures that refueling station personne) can be promptly informed of significant' changes in the facility status or core ceactivity conditions during CORE ALTERATIONS.

BYRON - UNITS 1 & 2 B 3/4 9-1

W G

4

- 9 ATTACHMENT B (continued) l BRAIDWOOD AFFECTED PAGES 4

1

\\

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS i

SECTION PAGE 3/4.0 APPLICABILITY...............................................

3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 1

3/4.1.1 BORATION CONTROL Shutdown Margin - T,yg > 200*F...........................

3/4 1-1 S hutdown Ma rgi n - T,yg, 200*F...........................

3/4 1-3 5

, Moderator Temperature Coef fi cient........................

3/4 1-4

'Mi nimus Temperature for Criticality......................

3/4 1-6 3/4.1.2 BORATION SYSTEMS F l ow P a t h - S h u t d own.....................................

3/4 1-7 Flow Paths - Operating...................................

3/4 1-8 Charging Pump - Shutdown.................................

3/4 1-9 Charging Pumps - Operating...............................

3/4 1-10 Borated Water Source - Shutdown..........................

3/4 1-11 Borated Water Sources - Operating........................

3/4 1-12

,J g

Boron Dilution Protection System.........................

3/4 1-13a l

3/4.1.3 MOVABLE CONTROL ASSEMBLIES j

Group Height.............................................

3/4 1-14 TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE v

EVENT OF AN INOPERABLE FULL-LENGTH R00..............

3/4 1-16 g

9 Position Indication Systems - Operating..................

3/4 1-17

'A Position Indication System - Shutdown....................

3/4 1-18 7

Rod Drop Time............................................

3/4 1-19 Shutdown Rod Insertion Limit.............................

3/4 1-20 Control Rod Insertion Limits.............................

3/4 1-21 FIGURE 3.1-1 ROD BANK INSERTION LIMITS VERSUS THERMAL POWER FOUR LOOP 0PERATION...........................

3/4 1-22

\\

FlGOE

3. i.- O Mown.AmoR, %wgAwg Cog p%p-1 LNL.

,..,,.S/tL-5 BRAIDWOOD - UNITS 1 & 2 IV Amendment No. g

~

g. pre Fgure J. d unk Qute 2F !

4

\\

J ow next Me.

6L 4

\\i

...:t. m..;.:

l ;

e j

... a.1..

f,.

.e h ".; =.U;;d., *~.;I _"."._ _'Ei_::_~_

~ "

_-~

. ;; J. -~.... :._:1..=. :_(-r y_ __.=..::::

g.,;;,

~ * - ~

=.3

• d. *;*"".,,.*. - [,;. - - -

,.f. ~ ' *

• ~

• .;;.,gn

=:.=5.

• ,,' E E ; 'a-

. g.4

< -. +

=,g,.

, m-s

.; g.

;;f.;;-;

...4 La- ". L.=.;. =. _.-__.._.u_._-._

- _. _ 660

..e--

.-,=. M..-

- - -.-w

,._g

=

-. -.- e.- --.

.=::

--e-

~7~ "I I

C C'.

...5 ~,;4,7^-

~~

_ - - ~,

~

n

=

....a,. + g.

., =

sm me..,

--=a==

y A

g:

T*r '- :

~;

_.......-.mg g =- -.=g...

6.th &

~

w

;-': ::. y ::.- :-

Cm r._

._::_:7

9=,-- a_.

~

.:.:.. c_

.:==.= ;

Zb5

.%2%n-L.-

. ^*-"

~hl'-

4 --~.

.%_L: =.=_=--3

,: = :

A.M

=.7__

=.,

-- =

". 640

. = -

---_... Q_&__-..

=_.

me m~a -..

.k

-+

.'-~ au. = *f=i%E4M:-4= 42=-s : % -

~

= = = _ -e_

w g

., ' ' ~ -

.- : u=.=-r_

. g,,,_

g;,

g

- =' ~ ~ -

=,. -

- - = -

==

v =-.2_ r N

c. _--

=

.y

= = =

-1s:. e--

w g

s:T::=_ :_

W k-

=-=.:.-

~

r,

...... l,......_

A

,=..

=.,;-

>=

- - x

._=:: =

_ _ =

t-

r

qi

.~;,, _ q...,...,. =,.

• .. - - -.:- M==...

m :-

3

,y

2;;;;.

y m.v-y

.:= %g,. g,_..

s W fSn " ".. *. -

[f-E"."."._-..=; : _.

. _, )

.r'; ;,.

".-. 6;;,.43... g.* f :.. _-

• As W&V

~- w 3

m-

'(.

~.1- ---

'-)- n - -.-.' 1-...

..'. ~ -'il 4" 4C 4;: Z

.u f

--. ' ". 4

...."I f..<.....

_....j

. =* J.J.;. =0.,x!..Lt u "a1

g.. :

+m-== 2 g.

wgs3

-- g y

==.

3 ::

- =. ".r}..

..........=-

w w:==:: =.= _..., ::,-..,.

_. -a :.w r-610

.._ _.. -- 3 4..

, m.. m g,

g d.

g.... -. _ --

c g :-,

.gs... -

_... ".'.. _ '

• _ =.

?** f

.: = T

-L

... ""5 $.._.g..

-a g

p.,

g

_' J

-* :"- --J:_

..1 g

..,._:,*.=::":=;.

..u.: =.*

-a

. :-. - - -: ft =

u--

.- =.. =. :.

%e...

= ----

,g

.==.-== -a..=

}.f-

-=::

.=...

[ f,,, y

j.N.... _.y-

\\

==.-

=-

- _..a

--- p

=.-- =.. --===.

=

. a_. =:._ =_=2_....sa.-...

590

~# ~=

=

~

~ ="=-~== =

= = 5 0 6 E

W_

=-*

1 =.=;.=

;=

==

_=..=. _ =.. =.._

A.

r= m U= =-

_- ;;;.=.; -- = :=.. =-__. -=2:-

_.._.W 3 s: 9 L1

==.

.=..-._ =. =.._,-- - -a

.= =

=

--/;==.

- ~ - ~ -

~ "- -

'=

....p........_

.".'.).,

.)

5

-1 d.

40 60 80 100 120 POWER (PERCEWT)

FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION 1

BRAIDWOOD - UNITS 1 & 2 2-2

n.

680 670 2471. psia.c... '. '..

e.

,,..,.3 3.,.

660 "2257 pain t r.-

r- - -..-

r.-. -

, r.- -

's.

650

_g m

640

s N.

2 0 0 0. P. s. i.a.....

~m u

CD

a. 630 E

.w

.s 1866 33..

\\. pg CD N.

\\...

cng 620

. g.

g 4

610 600 w

.,. ~...

N l

590

-. ~.....

580 7

1 1

f f

I f

f 1

f f

?

f 0

0.2 0.4 0.6 0.8 1

1.2 Power (Fraction of Nominal)

Figure 2.1-1 Reactor Core Safety Limit Four Loops in operation

-i kb\\ bV Q h *'*

/k ". hk i

_A A

u A

_....Aa u

4m-oa

+.

4-e -.- u A 4.-

.~

W 2.L SAFETY LINITS c

EM5 2.1.1 ttACTOR CDii The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation which would Msult in the release of fission products to the reactor coolant.

by restricting fuel operation to within the nucleate boiline regime w hest transfer coefficient is large and the claddin

' slightly above the coolant saturation tasperature.g surface temperature is Operation above the upper boundary of the nucleate boiling regime could result.in excessive claddin free nucleate bailing (DNB)g temperatures because of the enset of departure

-and the resultant sharp reduction in heat transfer coefficient.

DNS is not a directly measurable parameter during operation and-therefore THERNAL POWER and Reactor Coolant Temperature and Pressure have related to DNS.

This relation has been developed to predict the DNl flux and the location of DNS for axially unifom and nonunifom heat-flux distri-g' buttons.

The local DN8 heat flut ratio (DNBR) is defined as the ratio of the d

t heat flux that would cause DNB at a particular core location to the local heat

?

flux, and,is indicative of the margin to DN8.-

D b Th 8" d-Haa b".i

? ?? ee fe!!

. _A._htitt,u +6.+ eh..t #. nuam.,e r :

tha^ ="?* ** e* 1-et t'** =^---t

- _ _ _....... mm in-e*, - - > a a

>>i

!: ---der *

• er --_....'.i....21*-*',-","**...y.......,-....e--- ---

u'. =-' 1.

event:

. ~.

^

m...s r.u 5.a_2

_ _...._..... neia.m.

..e

u. --.-

..-...-.is -.......

fee! *- th'e==-"!?t i^=} _

T'a r

an+ 4. analrahla awnaefmantal datater -'-t he """* I S't ': =^ dt h t d t :-' =

o.

tk.

6 j amt auch that thaps 1.. at 5Fedf!!'; Mth ** -^- =t :="t=; ^Je,t :.2.M'i ; ; m.,

.;,5. E,. ii,I,2 4

i nume 4..... u... i.. e.. nun s u... i s...s..

,a s.. u-e^== i e t 4 --').

h r :tia'

• "a destaa k=? e, e^*edef =*8e! i= pla-t - :rMs.

i

- ch,;r,::: 'Ja =_m,1 ;;rr:t:::, =d '.=._1fd.t.h:th.;u;_;__r-~ ~ur; ;r;;; n, r.;'Or.;4 779.,

,=

.. _. _ _, < _ _.........,.-..<_..u __.................

.m_.

.m_

'e*

• At "etti" -^t is===ete= thia e r - -"- ! * ^ * " ""* * " -' *

?Mx:-

-t-Sti:: i-tr.: :h.; ; hat pr=; tere er; ;;d te Cter;,he 'Je i,1;nt -"

eace t&ty Na M8

-e+tebP:M; ; 4;;i;n ^"" ;;h: duh==t M =t ' phet =ht; :=1":Se^e==tet-ty, e:'= =h:: M ' M;;rr:u-- Mtheut==rS'-t'n.T The design DNIt values areD and respectivel for a ' typical cell and a thimble cell

'e r "" 'aa!, 2*' 1. ?? *^r 2 *y-kel et'. '

-' 1. ??

• r : *> 'd M 11 *~ *y.

z

"*"^a'

' '~ '

g,50 meeting safety analysis DNBF)1siu of'M4 for a typical cell and

)(

thimble es11 f.r "" '.;', = t.'? =d 1. hr : t;;h:1 =11 =d : ^J.idh

)

Isil, re;pcthely fer "J.; "T*C ", f.el in performing safety analyses.

1]

The curves of Figure 2.1-1 show the loci of points of THEANE POWER, Reactor Coolant Systen pressure and average temperature for which the etnisue design DNBR is no less than the desi gn DNBR value, or the average enthalpy at' '

the vessel exit is less than the entalpy of saturated liquid.

A 3r me _ $r y

A BRAIDWOOD UNITS 1 & 2 5 2-1 AmendmentNo.[

Insert A The DNBR thermal design criterion is the probability 'that DNB will not occur on the most limiting rod.and. is at ;least 954 (at a 95% confidence level) for any Condition I or II event.

In meeting this design basis, uncertainties in plant operating' parameters, nuclear and thermal parameters, and fuel fabrication parameters are considered.

As described in the UFSAR, the.

effects of these uncertainties have been statistically combined with the correlation uncertainty.

Design limit DNBR -values have-been determined that satisfy the DNB design criterion.-

/

i e

\\

TABLE 2.2-1 (Continued) 1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS k

7 FUNCTIONAL UNIT TRIP SETPOINT ALLOWA8LE VALUE i

] 12.

Reactor Coolant Flow-low k

>90% of loop mini-

>89.3% of loop

(

g mum measured flow

• minimum measured-

)

g flow

• e-13.

Steam Generator Water Level Low-Low m

a.

Unit 1

>33.0% of narrow

>31.0% of narrow span span

- ({

range instrument range instrument b.

Unit 2

>17%~(Cycle 3);

>16.3% (Cycle 3);

(

3 1 6.3% (Cycle 4

>34.8% (Cycle 4 and /

and after) of-after) of narrow i

m narrow range range instrument

(

instrument span span

14. Undervoltage - Reactor

>S268 volts -

>4920 volts -

Coolant Pumps each bus iach bus

(

'15.

Underfrequency - Reactor

>57.0 Hz

>56 08 Hz Coolant Pumps

(

16. Turbine Trip

,(

k(

a.

Emergency Trip Header 11000 psig

>815 psig Pressure b.

Turbine Throttia Valve'-

>1% open

>1% open

[

g Closure g17. Safety injection Input N.A..

M.A.

m-from ESF

18.. Reactor Coolant Puso N.A.

M.A.

5 Breaker Positlop. Trip k*Hinfeuemeasuredflow="

c gpm 3.?]SC'-

r

4 TABLE 2.2-1-(Continued) n.

b TABl.E NOTATIONS Ey NOTE 1: OVERTEMPERATURE AT h,

1. kk-iN5$(2 3 ydiAT.IKs-Keh--{^[ET(3 f g ) - T'] + Ks(P --P') - fs(a!)]

v Where:

AT Measured aT by RTD Manifold Instrumentation,

=

1[

lead-lag compensator on measured AT,

=

7 13, tr Time constants utilfred in lead-lag compensator for AT, t

=8s,

=

Tr = 3 s.

fg Lag compensator on measured AT,

=

1 7

w Time constants uttitzed in the lag compensator for AT, is = 0 s, ts

=

AT, Indicated aT at RATED THERMAL POWER,

=

1.164,*

- \\ hS "

K

=

0.0265/*F,*

O. cTAC\\~b / *T K,

=

f[

The function generated by the lead-lag compensator for T,,,

=

dynamic compensation,

[

Time constants utilized in the lead-lag compensator for T,,,,14 = 33 s,

14. Ys

=

Ts " 4 S.

g ff Average temperature,'*F, T

=

1 h

Lag compensator on measured T,,9,

=

1 5

f

)

hpM s e oMC

-b \\) d i _L a Q d %+

h owh\\ toq,WAiw c.Q 4Q V

w,- Mbeohk % Ay.i_ a.4 hs-1 SAncbg eth c(pO_

b.

y TABLE 2.2-1 (Continued)

TABLE NOTATIONS (Continued) 7 NOTE 1:

(Continued)

Time constant utilized in the measured T,yg lag compensator,.Is =. 0 s, t.

=

[.

T' 588.4*F.(Nominal T at RATED THERMAL POWER),

yg 0.00134,*

o.co W4 "

K

=

3 Pressurizer pressure, psig,.

P

=

2235 psig (Nominal RCS operating pressure),

P'

=

Laplace-trar + form operator, s 1, S

=

y and f (afi is a function of the indicated difference between top and bottom detectors of the t

power-ras.ge neutron ion chambers; with gains to be selected based on measured instrument co response during plant STARTUP lests,-s h t-G -wn-no wS, between -32 M nd +1 (aI) = 0, where.qt and q are Percent RATED THERMAL POWER (1) for qt 9

b b

in~the top and bottom halves of the core respectively, and qt *A is total THERMAL POWER in b

percent of RATED THERMAL POWER; i

exceeds 13% he aT Trip Setpoint shall be-(ii) for each percent that the magnitude of qt'9b

automatically reduced by 1.74% of its v4 at. RATED THERMAL R.

exceeds -32f, 2 -M'lEr,

+. u r" he AT trip setpoint.shall be k

(iii) for each percent that the. magni T q{- qb

.y automatically reduced by 1.6 of its valug t RATED THERMAL-POWER.

%Mt">

m.5 NOTE 2:

The channel's maximum Trip Setpoint snais not exceed its computed Trip Setpoint by more than g

3.71%*of-AT span.

L u>%

ma.

+ w w u 2 -a-ma sn s.

w-AW%tw % uua %

A wa6g uMqgh_ W

O TABLE 2.2-1 (Continued) h S

TABLE NOTATIONS (Continued) 8 NOTE 3:

(Continued)

{o.oo24s/ep

-5 0.oo170/*kor T > T" and K. = 0 for T <

T".

~

x.

=

T

=

As defined in Note 1,.

T"

= Indicated T, at RATED THERMAL POWER (Calibration temperature for AT instrumentation, < 588.4*F),

S As defined in Note 1, and

=

f (AI) 0 for all AI.

=

N The channel's maximum Trip Setpoint shall not exceed its computed Trip setpoint by more than

.g NOTE 4:

2.31%%f AT span.

30%

r 4,

\\(dkg N

g u % gt g OAN 1. ovv4 OA+ A hug wiE y 6.

m N

REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE C0 EFFICIENT l

LIMITING CONDITION FOR OPERATION 1

3.1.1.3 The moderator temperature coefficient (MTC) shall be:

a L"Less positive than 0 Ak/k/'F for the til rods withdrawn, hot zero THERMAL POWER condition, oe. cw-<i h wxt B b.

Less negative than -4.1 x 10 4 Ak/k/'F for the all rods withdrawn, and of cycle life (E0L), RATED THERMAL POWER condition.

APPLICABILITY: Specification 3.1.1.3a. ' MODES 1 and 2* only#.

Specification 3.1.1.3b. - MODES 1, 2, and 3 only#.

ACTION:

a.

With the MTC more positive than the limit of Specification 3.1.1.3a.

above, operation in MODES 1 and 2 may proceed provided:

1.

Control rod withdrawal limits are established and maintained sufficienttorestoretheMTCtolesspositivethan0ak/k/*F7 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. F(we Q.

These withdrawal limits shall be in addition to the insertion' limits of Specification 3.1.3.6; 4gk,.e g

2.

The control rods are maintained within the withdrawal limits established above until e subsequent calculation verifies that the MTC has been restored to within its limit for the all rods withdrawn condition; and 3.

A Specisi Report is prepared and submitted to the Commission pursuant to Specification 6.9.2 within 10 days, describing the value of the measured MTC, the interim control rod withdrawal limits, and the predicted average core burnup necessary for restoring the positive MTC to within its limit for the all rods withdrawn condition.

4 The provisions of Specification 3.0.4 are not applicable, b.

With the MTC more negative than the Ifmit of Specification 3.1.1.3b.

above, be in HOT SHUT 00WH within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

"With K,ff greater than or equal to 1.

1. See Special Test Exceptions Specification 3.10.3.

c 9@tte & Q y CA v A i_ M v w &vM %k4A Ay W-3 Ib*- bWbeo. k -\\c Cet 1 CLudL Uad A 6LCLdA v0c\\h c r, b,

BRAIDWOOD - UNITS 1 & 2 3/4 1-4 Amufdment No.

.=

l Insert B a.2."

Maintained within the limits specified in FIGURE 3.1-0, and

-W t

.ag_a.e 9

+

D

,, e -

6 10 9

[8 l Oncte.c pab\\t Ovaram uE f7 g

)6 ?

u

~

E 5

2 ro

-4 3

u S

$2 b

]1 3

0 O

10 20 30 40 50 60 70 80 90 100

" ;;;.- ' Pad:. Of '?:-t" Iir.we.w7 V<, \\ 9 pisovE 3 1 - O+#

%4eg Tugegh be%c sont %

e.b b OGI l

w hwk.oh\\t. 4 - Odfr 1 u_c\\ dA+ 3 %ct.c63 ui t+b Cp b-j u

i

/

k k

1.h hh\\h M D N\\ b

~:

m REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCE-- SHUTDOVN LIMITING CON 0! TION FOR OPERATION 3.1.2.5 As a minimum; one of the following borated water sources shall be OPERABLE:

a.

A Boric Acid Storage System with:

1)

A minimum contained borated water level of 7.0%,

2)

A minimum boron concentration of 7000 ppm, and 3)

A minimum solution temperature of 65'F.

b.

The refueling water storage tank (RWST) with:

1)

A minimum contained borated water level of 9.0%,

2k *A ministe hnenn enncantration of 20t)0 nnm amt b" b bcw whArv.%mk+weC5 $9D c%15t>D 3)

A minimum solution temperature of 35'F.

)Q APPLICABILITY: MODES 5 and 6.

ACTION:

I With no borated water source OPERABLE, suspend all operations involving CORE-ALTERATIONS or positive reactivity changes.

J SURVEILLANCE REOUIREMENTS 4.1.2.5 The above required berated water source shall be demonstrated OPERABLE:

a.

At least once per 7 days by:

1)

Verifying the boron concentration of the water, 1

2)

Verifying.the contained borated wate'r level, and j

3)

Verifying _the boric acid storage tank ~5elution temperat>>rt when-

'l

-it is the source of borated water.

]

b.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by' verifying tt;e RVST temperature when'it is'the source of borated water and the.outside air temperature is less than_35*F.

B e oh % di* L ouA '044 A w4\\ qqq,

+-

$9 5.

cr 8AAIDW000 - UNITS 1 & 2 3/4 1-11

%- M Rtodt.,h dtit i. a4 OO.k h S M g LM k 6.-

M s WEWT uo.

,.-.,r,-

REACTIVITY CONTROL SYSTEMS 80 RATED WAT,ER SOURCES = OPERATING LIMITING CONDIT!0w FOR OPERATION 3.1.2.6 As a siniaun, the following borated water source (s) shall be OPERABLE as required by Specification 3.1.2.2 for MODES 1, 2 and 3 and one of the following borated water sources shall be OPERA 8LE as required by Specification 3.1.2.1 for MODE 4:

l a.

A loric Acid Storage System with:

1)

A minimum contained borated water level of 40%,

2)

A minimum boron concentration of 7000 ppa, and 3)

A sinimum solution temperature of 65'F.

b.

The refueling water storage tank (RwST) with:

1)

A minimum contained borated water level of 89%,

2)()"A minimum boron concentration of 2000 p b

h bon cowtev.M*4h beWh 3)

A minimum solution temperature of 35'F, and 4)

A maximum solution temperature of 100*F.

APPLICABILITY: M00E5 1, 2, 3, and 4 ACTION:

With the Soric Acid Storage System inoperable and being used as one c' 4.

the above required borated water sources in MODE 1, 2, or 3, restore the system to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and borated to a SHUT 00VN MARGIN equivalent to at least 1% Ak/k at 200*F; restore the Soric Acid Storage System to OPERABLE status within the next 7 days or be in COLD SM.lTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />, b.

With the Rv5T inoperable in MODE 1, 2, or 3 restore the tad tc OPERA 8LE status within 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> or be in at least HOT STWBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTOOWN within the folic eing 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

c.

With no borated water source OPERABLE in MODE 4 restore one borated water source to OPERABLE status within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or be in COLD SHUT 00VN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

WUNA 50 & A 0 4 OLd.k 5 0 N N

%\\e%m 04 y s.

w _ A g u a 4o LLM i au-A % + 1 s M a BRAIDWOOO UNITS 1 &

'3/4 1-12 Ny W b MEk'.T 00,

.4 POWER 0!57R!8UT!0N LIMIT 8

~

3/4.2.3 RCS FLOW RATE AND NUCLEAR EWHALPY RI$f NOT CHMEL FACTOR LIMITING CONDITION FOR OPERATION Indi'ated Reactor Coolant System (RC$) total flow rata and Fh shall be 3.2.3 maintained as follows for four loop operation.

}-

376 VOC RCS Total Flowrata 1496;;400 spe, and a.

Fg" 31.55 [1.0 + 0.3 (1.0-P)] for 0FA fuel 7

b.

Fh 31.65 (1.0 + 0.3 (1.0-P)] for VANTAGE 5 fuel where:

Messured values of Fh are obtained by using the novable incore detectors. An appropriate uncertainty.of 4X (nominal) or greater shall then be applied to the measured value of [g before it is compared to the requirements, and THttMAL POWER _

,, RATED THERMAL POWER APPLICA811,171: MODE 1.

ACTION:

With RCS total flow rate or Fh outside the region.of acceptable operation:

)

Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> either:

a.

Restore RC5 total flow rata and U to within the above limits.

g 1.

or Reduce THERMAL POWER to less than 50K of RATED THERMAL PO 2.

and reduce the Power Range Neutron Flux-High Trip 5etpoint to less than or aquel to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

1 3/4 2-8 Amendment No.

3RA10wooo - UNITS 1 & 2 1

--e.

. m

d e

3/4.5 EMERGENCY CORE COOLING SYSTEMS 3 /4. 5.1 ACCUMULATORS LIMITING CONDITION FOR OPERATION 3.5.1 Each Reactor Coolant System accumulator shall be OPERABLE with:

a.

The isolation valve open and power removed, b.

A contained borated water level of between 31% and 63%,

A.

c.

A beror. concentration +(-between 1900 and 2100 ppm, and k h CMM24d4'Od:th h MD M.'%cC) powd c.

A nitrcge c:ve~ pressure of Detween 602 and 647 psig.

APP LIC A8 '.. i Y: CDE5 1, 2, and 3".

ACTION:

a.

With one accumulator inoperable, except as a result of a closed isolation valve, restore the inoperable accumulator to OPERABLE status within I hour or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b.

With one accumulator inoperable due to the isolation valve being closed, either immediately open the isolation valve or be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6' hours.

SURVEILLANCE REQUIREMENTS 4.5 1.1 Each accumulator shall be demonstrated OPERABLE:

a.

At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by:

1)

Verifying the contained borated water level and nitrogen k

cover pressure in the tanks, and 2)

Verifying that each accumulator isolation valve is open.

"Pressuruer pressure above 1000 psig.

k 4. g p h d t. % M t b Ow.4. C M A U M %.M CW oV y 5.

w _ Apq=b CAW _ 4t> OM i. cwd de* D_ sMs g odd uph_ b.

BRAIDWOOD - UNITS 1 & 2 3/4 S-1 AMENDMENT NO.

(

EMERGENCY CORE COOLING SYSTEMS 3/4.5.5 REFUELING WATER STORAGE TANK LIMITING CONDITION FOR OPERATION 3.5.5 The refueling water storage tank (RVST) and the heat traced portion of the RWST vent path shall be OPERABLE with:

a.

A minimum contained borated water level of 89%,

b.i A minimum boron concentration of 2000 ppe, "A h h habh 4%oo M M-DO %)

A minimum water temperature of 35'F, and c.

d.

A maximum water temperature of 100*F.

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

ACTION:

With the RWST inoperable, restore the tank to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.5.5 The RVST shall be demonstrated OPEPABLE:

a.

At least once per 7 days by:

1)

Verifying the contained borated water level in the tank, and 2)

Verifying the boron concentration of the water.

b.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying the RWST temperature when the outside air temperature is either less than 35'F or greater than 100*F, and c.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying the RWST vent path temperature to be greater than or equal to 35'F when the outside air temperature is less than 35'F.

• - @ ic h M M _1. M M 4 2 U M Ob ed4 6%

oV @ s'.

q, % U d \\e. k O a h 1. aN A & 1 S W% OW f b e BRAIDWD0D - UNITS 1 & 2 3/4 5-11 AMENDMENT NO J6'

f' f

3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 3.9.1 The boron concentration of all filled portions of the Reactor Coolant System and the refueling canal shall be maintained uniform and sufficient to ensure that the more restrictive of the following reactivity conditions-is met:

A K,ff of O M5 or less, or a.

b.t A boron concentration of greater than or equal to 2.000 ppm.

h bww Co*caMMW 09 C3tud:ct -tig.,sur-en gol AD '23CCppg.

O APPLICA ILITY: HODE 6*.

T ACTION:

With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity

r.:..g:: :..: i..i.i -.: cr.: c:.r.:ir.a: :,;,retier. ;; greeter than or equal to 30 gp.

of a solution containing greater than or equal to 7000 ppm boron or its equiv-alent until K is reduced to less than or equal to 0.95 or the boron-concentration is restored to greater than or equal to 2000 ppm %,c',r30o;>pM eff

( whichever is the more restrictive.

SURVEILLANCE REQUIREMENTS 4.9.1.1 The more restrictive of the above two reactivity conditions shall be determined prior to:

a.

Removing or unbolting the reactor vessel head, and b.

Withdrawal of any full-length control rod in excess of 57 steps (approximately 3 feet) f rom its fully inserted position within the reactor vessel.

4.9.1.2 The boron concentration of the Reactor Coolant System and the refueling canal shall be determined by chemical analysis at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

4.9.1.3 Valves CV1118, CV8428, CV8441, Cv8435, and CV8439 shall be verified closed and secured in position by mechanical stops or by removal of air or electrical power at least once per 31 days.

"The reactor shall be maintained in MODE 6 whenever fuel is in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.

AhepbcA4 lao Odk i cu.ui cwa a, um%\\ cogk.Btw oQ

% c.

BRAIDWOOD - UNITS 1 & 2 3/4

-1 g g ico.h 4,,, % O M i an d 4*h. M ] d

.b.

AvAEMbMtsu'T tJD,

i s

l 9

REACTIVITY CONTROL SYSTEMS I

BASES

~

MODERATOR TEMPERATURE COEFFICIENT (Continued)

The most negative MTC value equivalent to the most positive moderator density coefficient (MDC), was obtained by incrementally correcting the MDC These corrections used in the FSAR analyses to nominal operating conditions.

involved subtracting the incremental change in the MDC associated with a core condition of all rods inserted (most positive HDC) to an all rods withdrawn condition and, a conversion for the rate of change of moderator density with This value of the MDC was then temperature at RATED THERMAL POWER conditions.

The MTC transformed into the liatting MTC value -4.1 x 10 4 Ak/k/*F.

value of -3.2 x 10 ' Ak/k/*F represents a conservative value (with corrections for burnup and soluble boron) at a core condition of 300 ppe equilibrium boron concentration and is obtained by making these corrections to the limiting MTC value of -4.1 x 10 4 Ak/k/'F.

b The Surveillance Requirements for measurement of the MTC at the beginning

(

and near the and of the fuel cycle are adequate to confire that the MTC can be The BOL MTC sensurement combined with the maintained within its liatts.

predicted MTC with core burnup can be used to impose admin

(

l

)

inis coefficiant changes slowly due princt Ally _to thal *tductionmin RCLbh e\\('th J W

.ho.o s to" ".cw/v *% wedTeQ'-

concentration associated with fuel burn pm \\e w%

MINIMUN TEMPERATURE FOR CRITICALI _ % *Ng@Qkg 3/4.1.1.4 Qov inrmade crit 4 cal This specification ensuras that the reactor will av This with the Reactor Coolant Systes average temperature less than 550'F.

(1) the moderator temperature coefficient is limitation is required to ensura:

within its analyzed temperature range, (2) the trip instrumentation is within its normal operating range, (3) the pressurizer is capable of being in an OPERABLE status with a steam bubble, (4) the reactor vessel is above its temperature, and (5) the plant is above the cooldown steam dump minimue RTHDT permissive, P-12, 3/4.1.2 BORATION SYSTEMS The Boron Injection Syster ensures that negative reactivity control is The components required to available during each MODE of facility operation.(1) borated water sources, (2) perfore this function include:(3) separate flow paths. (4) boric acid transfe power supply from OPERABLE diesel generators.

With the RCS average temperature above 350'F, a sinimum of two boron the event an assumed failure renders one of the flow pa The boration capability of either flow path is sufficient to provide a SHUTDOWN MARGIN from expected operating conditions of 1.3% Ak/k after cooldown to 200*F.ct E0Ef-fel' peeer eg'!11 brit = renen ccnditien cnd rc:;utru (t3,%T) ~15 7804a11ons of 7000-ppe borated water from the b eeeur:

l

/0, gallons of 2000-p borate M>ot47 aato-p;m

-2 Amendment No. g

- o

+ g D = UNITS 1 &

w e a sww p s j

BRA

.p.- :..y.,

REACTIVITY CONTROL SYSTEMS BASES B0 RATION SYSTERS (Continued) WY A Boric Acid Storage Systee lev of 40% ensures that there is a volume of greater than or equal to 15,780 gallons available. A RWST level cf 89% ensures l

that there is a volume of greater than or equal to 395,000 gallons available.

With the RCS tesperature below 350*F, one Boron Injection System is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single Boron Injection System becomes inoperable.

The limitation for a maxists of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps except the required OPERABLE pump to be inoperable below 330'F provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORY or an RHR SuctiorF4alve.' re3,e.Q [g gpgw l

The boron capability required below 200*F i ufficient provide a SHUTDOWN MARGIN of 3% Ak/k after xenon decay d cooldo rom 200'F to f

140*F. This condition requires either 2,652 allo dif 7000 ppe borated / @@TPN water from the boric acid storage tanks or 11,840 allons of 2000 ppervborated water from the refueling water storage tank (RWST). A Boric Acid Storage System level of 7% ensures there is a volume of greater than or equal to 2652,ga11ons

[

available. An RWST level of 9% ensures there is a volume of greater than ar equal to 38,740 gallons available.

(y g)+

The contained water volume limits include allowance for water not available because of discharge line location and other physical characteristics.

B.

The limits on contained water and boron concentration of the RWST also ensure a pH value of between nd 11.0 for the solution recirculated within containment af ter a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.

The OPERABILITY of one Boron Injection System during REFUELING ensures that this system is available for reactivity control while in MODE 6.

The OPERABILITY of the automatic Boron Oilution Protection System ensures (7

adequate capability for negative reactivity insertion to prevent a transient caused by the uncontrolled dilution of the RCS in MODES 3,4, and 5.

The func-k tioning of the system precludes the necessity of operator action to prevent fur-

[

ther dilution by terminating flow to the charging pump (s) from possible unborated N

water sources and initiating flow from the RW5T.

The most restrictive condition occurs shortly af ter beginning of life when the critical boron concentration is highest, and a 205 gpe dilution flowrate provides the maximum positive reactivity addition rate.

One reactor coolant pump in operation with all reactor coolant loop stop isolation valves open reduces the reactivity addition rate by mixing the dilution through all four reactor coolant loops. A minimum count rate of I

ten counts per second minimizes the impact of the uncertainties associated with the source range nuclear instrumentation.

In the analysis of this accident, a minimum SHUT 00WN MARGIN of 1.3 Ak/k is required to control the reactivity tran-sient. Actions taken by the microprocessor if the neutron count rate is doubled will pravent return to criticality in these MODES.

b BRAIDWO0D - UNITS 1 & 2 B 3/4 1-3 Amendment No. f(

coh.CJL, % Mt i and Out RMM Wd Ptp@A b.

t W

7

~_

...i l

I

. ~..

POWER DISTRIBUTION LINITS BA5ts HEAT FLUX HOT CHANNEL FACTOR, and RCS FLOW RATE AND NUCLEAR ENTHALP HOT CHAlelEL FACTOR (Continued)

The control rod insertion limits of Specification 3.1.3.6 are c.

maintained,and The axial power distribution, expressed in teras of AXIAL FLUX d.

O!FriRENCE, is maintainsd within the limits.

Fh will be maintained within its limits provided the Conditions a.

y The combination of the AC5 flow requirement (

.gpe)

{

d. above are saintained.N guarantee that the DNBR used in the safety analysis and the requirement on Fg will be set.

-p

-[

..s u...,-..a,Rs,( 1. 'O : d 1. 0 f: r t h O'*.

l is limit DN8 Margi ft m,,,, s.a a.,between the sa e,,y ana ys 955bilEypical and thimble cellsf and the dedigEiiii DUIt g,7 gg, a,

..a e u,..,

~

( _ e f:r th: 0" fd tn': ! :4 tMd'^ ^1?:, : d 1.?? : t 1.?? for the r

f[>

in;::t';;?;) is maintained.

-VAE'S 5 fee' typical and thiable cellsx A fraction of this margia is utilized to acceanodate the trene464eMere riate fuel red bow ONBA penalty

(

""'"" gx",, 'r:'::: :f 1"..".".) ;c4 t!.; a p he rest of the mergin between

'(

(less than 1.5 per WCAP 8691 Revision 1.

design and safety analysis DNBR limits,can be used for plant design flexibility.*'

52emeh V

QQ,b50 The RCS flow requirement is based on thdloop sinfaum asasured I:

I of b spe'which is used in the Mr:::f hermal Design Precedure,desselbed,-

j-

h. T*.;." 4.4.1.;4 3.4:t A precision heat balance is performed esos each cycle Potential fauling of the and is used to calibrote the RC$ flow rate indicatore.

feedvater venturi, which might not be detected, could bias the resulte from the-a penalty of precision heat balance in a non conservative manner..Therefore,imum seasurement I-0.1% is assessed,for potential feedwater venturi fouling. A anx p

uncertainty of eff has been included in'the loop sintaus asasured flav rate to account for potential undetected feedwater venturi fouling and the use of the RCS flow indicators for flow rate verification. Any fooling which might bias l

the RCS flow rate 'neasurement creater than 0.1% can be detected by monitoring If detected, action shal be taken, before performing subsequent precision heat balance m and trending various plant performance parameters.

l either the effect of fouling shall be quantified and compensated for in the RCS-flow rate seasurement, or the venturi shall be cleaned to eliminate the fouling.

Surveillance Requirement 4.2.3.4 provides adequate monitoring to detect possible flow reductions due to any rapid core crud buildup. -

Surveillance Requirement 4.2.3.5 specifies that the sensurement instrumen-tation shall be calibrated within seven days prior to the performanca of the-i This requirement is due to the fact that the l.

calorimetric flow asasurement.

drift effects of this instrumentation are not included in the flow asas This requirement does not apply for the instrumentation uncertainty analysis.

whose drift effects have been included in the uncertainty analysis.

(

j AmendmentNo.f BRAIDWOOO - UNITS 1 & 2 8 3/4 2-4 s

uj

=

a,. :.

EMERGEMCY CORE COOLING SYSTEMS BASES 3/4.5.5 REFUELING WATER STORAGE TANK The_ OPERABILITY of the refueling water storage tank (RWST) as part of the ECCS ensures that a sufficient supply of borated water is available for injection by the ECCS in the event of a LOCA.

The limits on RWST minimum volume and baron concentration ensure that: (1) sufficient water is available within containment to permit recirculation cooling flow to the core, and (2) the reactor will remain subcritical in the cold condition following sixing of the RWST and the RCS water volumes with all control rods inserted except for the most reactive control assembly.

These assumptions are consistent with the LOCA analyses.

The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.

A minimue contained borated water level of 89% ensures a volume of greater than or equal to 395,000 gallons.

g The limits on contained water y also ensure a pH value of between use and boron concentration of the RVST nd 11.0 for the solution recirculated within containment after a LOCA.

This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.

BRAIDWOOD - UNITS I & 2 8 3/4 5-4 AMENDMENT NO.

4 -

1 CONTAIMENT SYSTEMS 1

BASES

~

CONTAINMENT PURGE VENTILATION SYSTEM (Continued) be exceeded in the event of an accident during containment purging operation.

Operation with one line open will be liatted to 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> during a calendar e '-

year.

Leakage integrity tests with a maximum allowable leakage rate for containment purge supply and exhaust supply valves will provide early indica-tion of resilient material seal degradation and will allow opportunity for repair before gross leakage failures could develop.

The 0.60 L leakage limit of Specification 3.6.1.2.b. shall not be exceeded when the leak $ge rates determined by the leakage integrity tests of these valves are added to the previously determined total for all valves and penetrations subject to Type 8 and C tests.

3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS 3/4.6.2.1 CONTAINMENT SPRAY SYSTEM The OPERABILITY of the Containment Spray %ytta= ansures that containent cepressurt2ation and cooling capability will be available in the event of a LOCA or steam line break.

The pressure reduction and resultant lower containment leakage rate-are consistent with the assumptions used in the safety analyses.

The Containment Spray System and the Containment Cooling Systes are redundant to each other in providing post-accident cooling of the containment atmosphere.

However, the Containment Spray System also provides a mechanism for removing iodine from the containment atmosphere and therefore the time requirements for restoring an inoperable Spray System to OPERABLE status have been saintained consistent with that assigned other inoperable ESF equipment.

3/4.6.2.2 SPRAY ADDITIVE SYSTEM O.D The OPERABILITY of the Spray Additive System ensures that sufficient NaOH l

is added to the containment spray in the event of a LOCA.

The limits on HaOH volume and concentration ensure a pH value of between and 11.0 for the solution recirculated within containment after a LOCA.

This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components. The contained solution volume limit includes an allowance for solution not usable because of tank discharge line location or other physical characteristics.

These assumptions are consistent with the iodine removal efficiency assumed in the safety analyses.

A spray additive tank level of between 78.6% and 90.3% ensures a volume of greater than or equal to 4000 gallons but less than or equal to 4540 gallons.

BRAIDWOOD - UNITS 1 & 2 B 3/4 6-3

@W b M EtrT Nio.

t 3/4.9 ' REFUELING OPERATIONS 4

BASES 3/4.9.1 BORON CONCENTRATION The limitations on reactivity conditions during REFUELING ensure that:

(1) the reactor will remain subcritical during CORE ALTERATIONS, and (2) a-uniform boron concentration is maintained for reactivity control in the water volume having direct access to the reactor vessel. The limitation on Keff of no greater than 0.95 is sufficient to prevent reactor criticality during

' \\+ -

refueling operations and includes a 1% Ak/k conservative allowance for fbh D) uncertainties.

Similarly, the boron concentration value of 2000 ppm %r greater includes a conservative uncertainty allowance of 50 ppm. These limitations are consistent with the initial conditions assumed for the boron dilution incident in the safety analyses.

The locking closed of the required valves during refueling operations precludes the possibility of uncontrolled' boron dilution of the filled portions of the RC5.

This action prevents flow to the-RCS of unborated water by closing flow paths from sources of unborated water.

3/4.9.2 INSTRUMENTATION The OPERABILITY of the Source Range Neutron Flux Monitors ensures.that redundant monitoring capability is available to detect changes in the reartivity conoition of tne core.

_3/4.9.3 DECAY TIME The minimum requirement for reactor subcriticality prior to movement of irradiated fuel assemblies in the reactor vessel ensures that sufficient time has elapsed to allow the radioactive decay of the short-lived fission products.

This decay time is consistent with the assumptions used in the safety analyses.

3/4.9.4 CONTAINMENT BUILDING PENETRATIONS Tne requirements on containment building penetration closure and OPERABILITY ensure that a release of radioactive material within containment will be restricted f rom leakage to the environment. The OPERABILITY and closure restrictions are sufficient to restrict radioactive material release from a fuel element rupture based uoon the lack of containment pressurization potential while in the REFUELING MODE.

)

The Braidwood Station is designed such that the containment opens into the l

fuel building through the personnel hatch or equipment hatch.

In the event of a fuel drop accident in the containment, any gaseous radioactivity escaping from the' containment building will be filtered through the Fuel Handling Building-Exhaust Ventilation System.

3/4.9.5 COMMUNICATIONS The requirement for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the facility status or core reactivity conditions during CORE ALTERATIONS.

9yyBcokh % Odck 1 Owr 1 t A "a b a b

. \\ciN y

t BRAI U ITS I & 2 B 3/4 9-1 pamwen m.

e e

ATTACHMENT 3 EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATION Commonwealth Edison (CECO) has evaluated the proposed amendment and determined that it involves no significant hazards consideration. According to 10 CFR 50.92(c), a proposed amendment to an operating license involves no significant hazards if operation of the facility in accordance with the proposed amendment would not:

(1)

Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2)

Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3)

Involve a significant reduction in a margin of safety.

The proposed changes would modify the Technical Specifications concerning (1) the moderator temperature coefficient (MTC), (2) the boron concentration necessary to meet shutdown margin (SDM) requirements, and (3) the thermal design flowrate.

The MTC change would allow a slightly positive MTC_(PMTC) below 100 percent of rated full power. The principal benefit of this change is that it would facilitate the design of future reload fuel cycles. Technical Specification changes are also required to meet SDM requirements to accommodate the positive MTC and the potential of lengthened reload fuel cycles due to increased energy requirements. To assure subcriticality requirements are met following a postulated loss-of-coolant accident (LOCA), the boron concentration is increased for the refueling water storage tank (RWST) and the accumulators. The safety analyses for the Byron and Braidwood Updated Final Safety Analysis Report (UFSAR) transients have been previously based on a maximum MTC being less than or equal to O pcm/ F at all times when the reactor is critical. The proposed change to the Technical Specification would allow a +7-pcm/*F MTC for power levels up to 70 percent with a linear ramp to O pcm/*F at 100 percent power. CECO has reviewed the revised USFAR safety analyses which conservatively bounds this positive MTC, increase in boron concentration, incorporates revised thermal design flows, and addresses increased tube plugging levels. The results of the revised analyses are provided in-WCAP 13964 " Commonwealth Edison Company Byron and Braldwood Units 1 and 2 -

Increased SGTP/ Reduced TDF/PMTC Analysis Program Engineering / Licensing Report".

l The thermal design flow (TDF) is a minimum RCS flow value assumed in the accident analyses and reactor core thermal / hydraulic design calculations that demonstrate the necessary heat removal from the core to meet various transient acceptance criteria. The minimum measured flow (MMF) currently used for the licensing basis is a total core flow of 390,400 gpm for Byron /Braidwood and is l

r l

r

~

i reflected in Technical Specification Table 2.2-1 (Functional Unit 12) as a footnote of 97,600 gpm per loop for the reactor coolant flow-low reactor trip. The MMF value i

must be verified in accordance with Technical Specification 3/4.2.3.

A reduction in TDF has been factored into the accident analyses that rely on RCS flowrate. This results in a reduction in the limiting condition for operation (LCO) value for RCS flow reflected in the Technical Specifications. The reduced flow requirement provides a margin to account for steam generator tube plugging (SGTP).

The revised TDF value corresponds to a MMF value of 371,400 gpm, which assumes a 3.5 percent flow measurement allowance, and is reflected in the footnote to Technical Specification Table 2.2-1 as 92,850 gpm, minimum measured loop flow for the reactor coolant flow-low reactor trip. The revised LCO flow value shall be incorporated in Technical Specification 3/4.2.3.

The proposed changes also include an administrative change to correct the wording in the MTC Technical Specification LCO 3.1.1.3a to clarify that both LCO 3.1.1.3a and b must be met over the fuel cycle.

Based on Commonwealth Edison's review and approval of WCAP 13964, which used NRC approved safety analysis methodology provided by Westinghouse and Commonwealth Edison, it has been determined that the changes associated with the analyses do not involve a significant hazard. Specifically:

A.

The proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

(1) Tne reduced thermal design flow and positive moderator temperature coefficient program, which includes corresponding changes to the RWST and accumulator required boron concentration, will not affect the operability and integrity of plant systems and components. The analysis program does not result in a condition where the design', material, and construction standards that were applicable prior to application of the program are altered. Additionally, the safety functions of the evaluated systems and components have not changed. The safety analyses necessary to support the reduced TDF and PMTC program were performed (WCAP 13964) and found to be acceptable and consistent with the Byron and Braidwood original safety analysis bases. All Departure from Nucleate Boiling (DNB) Ratio (DNBR) design limits were determined such that there was a 95 percent probability at a 95 percent confidence level that a DNBR value of 1.25 for a typical and thimble cell were verified to have been met. The present Technical Specification limit for Nuclear Enthalpy Rise Hot Channel Factor, F%, of less than 1.65 ensures that the limiting DNB ratio during normal operations and operational transients (Condition I and Condition 11 events) is greater than 2

r.

or equal to the DNBR limit of the correlation being applied.

The accidents'which are found to be sensitive to PMTC were analyzed as part of this effort and the results were found to be acceptable. On a cycle-by-cycle basis, the impact of PMTC on Anticipated Trip Without Scram (ATWS) risk will be addressed by determining the Unfavorable Exposure Time (UET) per established Westinghouse Owners Group methodology, with corrective actions to be taken as appropriate to assure acceptable risk. The increase in the RWST and accumulator boron concentration will have no adverse impact on the previously evaluated accidents. The SGTP/TDF/PMTC program does not affect the integrity of the safety related systems and components such that their function to control radiological consequences is affected and all fission barriers will remain intact. The effects on offsite doses have been considered. The incorporation of a PMTC, a reduction in TDF and increased tube plugging levels have increased offsite doses. However, the increases are small and the total doses are a small fraction of the 10 CFR 100 limits. As such, the acceptance criteria continue to be satisfied.

Therefore, the probability or consequences of an accident previously analyzed in the UFSAR is not increased by the SGTP/TDF/PMTC program.

B.

The proposed changes do not create the possibility of a new or different type of accident from any accident previously evaluated.

(2) The methodology and manner of plant operation as a result of the 3

proposed changes is unchanged. The increased SGTP, reduced TDF, and PMTC program, which includes changes to the RWST and accumulator boron concentration, does not impact the safe operation of the reactor provided that the existing and proposed Limiting Conditions for Operation (LCOs) and the associated action requirements are satisfied. The assumptions do not. create failure modes that could adversely impact safety related equipment. The related Safety Limits and LCOs in the plant Technical Specifications will be addressed and evaluated for each reload core design via the 10 CFR 50.59 process. All DNBR design limits were determined such that there was a 95 percent probability at a 95 percent confidence level that a design DNBR value of 1.25 for a typical and thimble cell were verified to have been met. Other than the analysis for tube plugging, the proposed changes do not involve any equipment additions or modifications at the stations. Currently installed equipment will not be operated in a manner different than I

previously operated. Changes will be made to technical data within the existing station procedures, however, the analytical methods used to determine the data will remain unchanged. All aspects of the 3

o a

SGTP/TDF/PMTC program have been evaluated, and no new or different

- accidents or failure modes have been identified for.any system or component important to safety. Also, no new credible limiting single failure has been created. Because the SGTP/TDF/PMTC program does not adversely affect the integrity of the steam generator or any other equipment, it is determined that an ' accident different than any evaluated in the UFSAR will not be created.

C.

The proposed changes do not involve a significant reduction in a margin of safety.

(3)

The performance and integrity of the evaluated safety-related systems and components are not affected such that their control of radiological consequences is altered. The reduced TDF and PMTC program, which includes changes to the RWST and Accumulator boron concentration, will have no effect on the availability, operability, or performance of the evaluated safety-related systems or components. The margin of safety associated with the licensing basis safety analysis is not reduced by the changes. All acceptance criteria for the specific UFSAR Chapter 15 safety analyses (Non-LOCA and LOCA) have been either evaluated or verified to be met using NRC approved methodologies. Therefore, there is no significant reduction in the margin of safety as defined in the

~

bases to any Technical Specification.

Based on the above evaluation, Commonwealth Edison has concluded that implementation of a PMTC, revised RWST and accumulator boron concentrations, and reduced RCS thermal design flow does.not involve significant hazards consideration with respect to the provisions of 10CFR50.92.'

i

-l l

4

' ;1 4

l ATTACHMENT 4 ENVIRONMENTAL ASSESSMENT Commonwealth Edison has evaluated the proposed changes associated with the SGTP/TDF/PMTC program against the criteria for and identification of licensing and regulatory actions requiring environmental assessment in accordance with 10CFR51.21. It has been determined that the proposed changes meet the eligibility -

criteria for categorical exclusion set forth 10CFR51.22(c)(9). This determination is based on the fact that this change is being proposed as an amendment to a license issued pursuant to 10CFR50, it involves changes to a surveillance requirement and the amendment meets the following specific criteria:

(i) the amendment involves no significant hazards consideration.

As demonstrated in Attachment 3, this proposed amendment does not involve any significant hazards considerations.

(ii) there is no significant change in the types or significant increase in the amounts of any effluents that may be released offiste.

The effects on offsite doses have been considered. The incorporation of a PMTC, a reduction in TDF and increased tube plugging levels have increased offsite doses. However, the increases are small and the total doses are a small fraction of the 10 CFR 100 limits. As such, the acceptance criteria continue to be satisfied. The proposed program assumptions do not change, degrade, or prevent the response of the evaluated safety-related systems and components such that their function in the control of radiological consequences is affected.

(iii) there is no significant increase in individual or cumulative occupational radiation exposure.

This proposed change will not result in changes in the operation or configuration of the facility; there will be no change in the lovel of controls or methodology used for processing of radioactive effluents of handling of solid radioactive waste nor will the proposal result in any change in the normal radiation levels within the plant. Therefore, there will be no increase in individual or cumulative occupational radiation exposure resulting from this change.

Commonwealth Edison has evaluated the proposed amendment against the criteria and found the changes meet the categorical exclusion permitted by 10CFR51.22(c)(9).

(:;.w.,-

J]

ATTACHMENT 5 i

WCAP 13964 Revision 1 Commonwealth Edison Company Byron and Braidwood Units 1 and 2 increased SGTP/ Reduced TDF/PMTC Analysis Program Engineering / Licensing Report 1

)