Text

Proposed,Reformatted Tech Specs Identifying Changes to Make Specs Applicable to Both Units
	ML20086U414
Person / Time
Site:	Comanche Peak
Issue date:	01/02/1992
From:	; TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC)
To:	;
Shared Package
ML20086U406	List: ML20086U403; ML20086U414;
References
	NUDOCS 9201080011
	Download: ML20086U414 (127)
	v • d • e

Txx-92001 ATTACHMENT 1 PAGE 2 Or 3

!NDEx 1AFETY LIMITS AND' LIMITING SAFETY SYSTEM SETTINGS SEC'!ON D_A G E.

21 SAFE'/ LIMITS 2.1.1 REACTOR CORE.

21

.2: 1. 2 RE Aticil, CCO L ANT S Y STEM P R ESSUR E...........................

x 2-1 FIGURE 2.1(lagfACTORCORESAFETYLIMIT............................

DI 3-Of4LT 11 2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS...............

2-3 TABLE 2.2 1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS.

2-4 BASES SECT:CN DAGE 2.1 SAFETY LIMITS 2.1.1 REACTOR CORE.....

B 2-1 L

2.1.2 REACTOR COOLANT SYSTEM PRESSURE...............

B 2-2

2. 2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP0!NTS...........

B 2-3 FIGURE 2.l-lb UNIT 2 REACTOR CORE SAFETY LIMIT.....

2-Za

(

llMANO:-E :E As

'.iN!T 1 iii i

SA10:oo3:

py g l

' FUR /IDOCF: G5d,oM, ; n,

[

F D C.

1 TXX-92001 ATTACHMENT 1 PAGE 3 0F 3 INCE(

AtulNISTRAT!vE : NT43L5

,r

( g a t * *- N 4..

3-a.,ut r.: c e N.: t = '. '.,.

6.2 :: Sani: N.

6-1

-v...4 ry.:, :

s.: :

e.

o.+4st 4;4r.y.

2.

a a

s iNl....-.

EWO UdfTc WIT +4 A CCNt40N CouTROL RCCth.

4 o. g. s s 2

~r*

r

{.3 TABLE 6.2 1 MINIMLM 5HIFT CREW CCMPOSIT:CN If#gl'hD//H[QA 6-3 6.2.3 INDEDENDENT SAFEIY ENGINEERING CRQVP (13EO)

Function.

6-4 ConDosition.

6-4 Responsittlities.

6-4 Recorcs....

6-4 6.2.4 SHIFT TECHNICAL ADVISOR.

6-4

6. 3 UNIT STAFF CUALIF! ATIONS.

6-4 6.4 TRAINING.

55 6.5 REVIEW AND AUDIT.

65 6.5.1_ STATIONS OPERATIONS REVIEW CCMMITTEE ($0RC)

Function.

6-5 Composition.

6-5 Alternates....

65 Meeting frequency.

6-6

$uorum.

6*6 t

Responsioilities....

6-6 Recorcs.............

6-3 4

6.5.2 CPERAT!CNS REVIEW COMMITTEE (ORC)

Function.......,........

6-6 Composition..........

6-5 Alternates........................

6-9 Consultants =....,........

6-9 Meeting Frequency.

6-9 Quorum.,..

6-9 Review...........

6-9 Audits.

5-13 Records......

6-11 CCMANCHE EEAA - UNII 1 xv i

1 TXX 92001 ATTACHMENT 2 i

PAGE 1 0F S SECTION: 2.1

1TLH, PAGE fl JUSTIFICATION 2A 22 The WRB 1 DNB correlation and tSe Westinghouse 2-2a*

Improved Thermal Design Procedure (ITDP) is used for Unit 2,

while.the W3 DNB correlation and the t

Westinghouse Standard Thermal Design Procedure (STDP) is used for Unit 1 in the riesign of the core, The.

correlation and design procedure differences result in the reactor core safety limits dif f ering between the i

units.

2B B 21-The WRB 1 DNB correlation and the !TDP with a DNBR safety 1

analysis limit value of 1,49 is used for Unit 2.

The-W-3 DNB correlation and the STDP with a limit value of 1.30 is used for Unit 1.

The wording of the DNB BASES has been changed for the discussion to apply equivalently to both units.

20 B 2-1 The nuclear enthalpy rise hot channel f actor limit at RATED THERHAL POWER and-the power f actor multiplier for the nuclear enthalpy hot channel factor may differ between the two units (Unit.2 cycle 1 differs from the current Unit 1 parameters).

Therefore the nuclear

-enthalpy rise hot channel _ f actor equation is being written as a generic equation with the unit / cycle specific parameters being relocated to the CORE OPERATING LIMITS REPORT.

i I

l-j

Proposed New Page r

l L

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

~ ~ " ~ " ~ ~ ~

Txx-92001 ATTACHMENT 2 PAGE 2 OF 5 d

I N

t i

h.;ge-

~-..a es22E5 *S:3

o e E X *' * - *~,- '

N,% ',,

00.

s2235 ASIS l

0.

i w

2 n

s 965 #5!3 I

_~

\\m l

.c.

se1845 *S!,a l

~.

}

i l

T

$275! :.

i t

t

==-

3 l

=-. s t

t i

=--

l I

==-

~~~

.n c.

w cv

=-

c.

= :..

. c-

=. ~. ~ u = = v :.

~. ~ :

FIGURE 2.1 la.

(O 2~k ;

( Unit 1 REACT *R CORE 5AFETY LIMITS

=

, m.._,-

COMANCME DEAK : UNIT 1 2-2

TXX.92001 l

ATTACHMENT 2 PAGE 3 0F 5 t

i h'R, n ",-

5, J A. (

4 W-

.. / / / / / / / / T /

~~

2: U / / / / / '/

/. /

~Jm../m / / somgW

= ~/ M N

///

'Wm/. /MX

/; /

~

e /

V NN /

f u

/ / M*NJ NX / /

' / / / / M AN

~

1

~ / / *nV+ /: / / MN' '/

/ / / / / ' / / /./ \\N

~

~~

1

~ ' / / / / / / : / /

/N

' / /

/./

/> /

/ / / /

1; / /; / / / / / / / /

=-

2:

e e

j =

f z:

=g=.z

r r=-

=v.

=

1=

.1.g 03 g _,, _,,,E1 _ s

, k,.

_.m _ 1

_f

+

~

..-9 r

p w

a y

,-r-<=,

,--w--...,s--

,,---r

--r

.-s

TXX 92001 ATTACHf'E"T 2 PAGE 4 Or 5 ;

9;;,,:v Ba5ES 2.1 '

E I e est ':t*; i :' inis 53'etj ui-it Orevent. nee eat g :f,e'e' a:

0:ssi 'e :'a: ' g :er' rat'on ani:

t 9 aould result in tae release :' <*sst:n Or: ducts to tne cea:::e ::olant.

Over" eating cf the fuel 1 3ccir; 5 :re-4 ente: ty eestrictia; 'uel :ceration tg.itnin t*e oveleste ::ili g reg'?e

=rere the meat trans'ee ::ei:ieat is large and the cla00'ng s rfa:e te* eri-ture 15 slightly aco e tne coolant satsration temperature.

Operation atose t*e upper bouncery of the nucleate boiling regi*e could dqhresultinexcessiveclaccingterceraturesbecauseofthesssetofcecarta coefficient. fih 1 0 al "<B p at luy/ragio p NB.) i ce'nec/as

e

'sti /:'

1 j$

tnp/neag fl t at cul

afse

B

't par ic or ore oc tic to ne

cal hp'a t fl u x nd s4 ci' tiv of tn tar in.o r 48, DNB is et a di ett y j

[c o/ ant tempera ure arc re sur ealvrab pp am.er,uri g er 10 an th ref re T'.ERP P wER no ea:*.c na e 0 en ela ed DN R +trov. t..-

cofrel ti f.

ne

-3 NB orr la on as ee dev op t cre i:t.re jNB

[d i s,t r i b)/t i 0ps.]

us /and/be

.c c i. i o O' A9 /Or axi iv unif-rm nd onutifor. nelt #%

j 2,- 6

/

ahe ium al o' tr CNjd d-in ste dy- (at 00 atiyh,r/r3'

/

oce/atig/in se/.5, j

nal.ra and art /cip ted,ran 1en s i i

te to 1/30 S%

4 4 ue ore too Os o a 9P. cr;4abi ity at 95% con 1:e ce el 4 t

/

fe/n.neNr.%'

ill ar. i enc,en/as i a reo at mar in 0

^on

'edt.ct i r,p f

11 e

d i

Relow which the calculated ut@R is nole.v than the safety ady The :grves of Ugurell'1"show the loci ~of points'of THERMANERJ M 6

~

Reactor C:olant System cressure and average temp f-MM-i 5 eo 'eo: tN-1.3h or tFe average enthalpy at the vessel e it. i 5 wah f40 the enthalpy of saturate 4 liquid, l l,3 3 tb n

/

These curves are Dased on a nuclear enthalpy rise not channel 'actcr, ;N A f 1,*0^and a refererce m W ith : peek M 1. 5"Maxial power sna:e, d N

allowance is included for an increase in F at reduced po.er cased or tre 2-C exeression:

3H

= <E 5.1 + s. 21-P)i, 5

M)Whe,.eD af:tactte/

\\

1*

n of RATE THE.MA POW M b M'* "

j;WHg*ihese rest fiva

'"itin[thanthosecalculatesfar conditions aretor:

the range of all control rods fully withdrawn to the maximum allowable control red insertion assuming the axial power imbalance is within the limits of the f

(1.1) function of t*e OvertemDerature N-16 trip.

When the axial cower t

imbalance is not witnin tre tolerance, the axial power imba. lance effect en the Overtemperature N-15 trios.411 reduce the Setpoints to provide crotect on i

consistent with core Safety Limits.

~

COMANCHE DEAN - UNIT 1 B 2-1

t i

i TXX 90001 ATTACHMENT C PAGE 5 0F-5 t

f i

INSERT V i

DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to ONB.. This relation has been developed to predict the DNB flux and the location of DNB for axially uniform and non uniform h it flux distributions.

The local DNB heat flux ratio (DNBR). defined as the ratio of the heat flut that would cause DNB at a particular core location to the local heat flux, is 1

indicative of the margin to DNB.

The DNB design basis is that there must be at least a 95 percent probablitty that the minimum DNBR of the limiting rod during Condition I and 11 events is greater than or equal to the DNBR limit of the DNB correlation being used.

The correlation DNBR limit is established based on the entire applicable experimental data set such that there is a 95 percent probability with 95 percent confidence level that DNB will not occur when the minimum DNBR is at the DNBR limit, In meeting this design basis, uncertainties in plant l

operating parameters, nuclear and thermal parameters and fuel f abrication parameters are considered such that there is at least a 95 percent probability with 95 percent confidence level that the minirnum DNBR for the limiting rod is greater than or equal to the DNBR limit, in addition, margin has been maintained in the design by meeting saf ety analysis DNBR limits in performing safety analyses.

INSERT y He F

F (1.0 + PF (1.0 P)]

3M P

~

8M where:

P

-the fraction of RATED THERHAL POWER (RTP),

F "' - the F', limit-at RTP speciliedintheCOREOPERATINGL1HITSREPORT(COLR),

and the power factor multiplier for F' PF specified in the COLR.

l 1.

l

7 TXX 92D01 ATTACHMi,NT 3 PAGE 1 0F 8 SECTION: 2.0 and BASES 2.2 M

PAGE 4 JUSilflCATIQR

- A 24 A2 The Unit 1 overtemDerature N 16 IOTN-It.)tetpoint is ba',ed 27 on core safety limits developed with the W3 DNB t

2-8 correlation and the Westinghouse Standard Thermal Design i

Procedure (51DP).

Unit 2 OTN 16 setpoint equation is i

based on core saf ety limits developed with the WRB 1 DNB correlation and the Westinghouse Improved Thermal Design Procedure (ITDP).

Through the use of the WRB 1 DNB l

correlation and the ITDP. significant improvements in margin a _r e realized i n the OTN+16 equations, which results in an improvement in the 'TA' and 'Z' terms.

The R2 and K3 values in the OTN 16 setroint equations,

}

which dif fer for the two units because of the dif f ering core safety limits. have an impact on the sensor

'S'

- terms.

1 Because of the significant margin available in the Unit 2 OTN 1fs setpoint, a provision has been made in the Statistical Setpoint Study to allov the use of the plant l

computers in the normalization of the N 16 power monitor indication to the results of the daily power colorimetric. This accounts f or the additional term in the OTN 16 and overpressure H + 16 ' S ' terms.

The sensitivities of - the N 16 power signal to changes i n temperature and pressure have also been developed in a more rigorous manner during the Statistical Setpoint Study, for Unit 2 than for Unit 1.

Additionally, the Unit 2 calculation includes a specific allowance for the i

cold leg streaming phenomenon-identified in Unit 1.

3B 24 A2 Unit 1 uses Barton pressure transmitters.

Rosemount 25 A2 pressure transmitters are used in Unit 2.

Changes in

[

the ambient temperature and pressure _ have slightly L

greater ef f ects _ on the flotemount t ransmitters t hen on r

the Barton transmitters.

Therefore, the 'Z' term will be slightly_ larger.for Unit 2 than for Unit 1.

1 A long term WQative drif t phenomenon has been identified which-af fects most Barton pressure transmitters.

The 4

Rosemounts are unaf f ected. Theref ore, the

'2' term f or

+

the pressurizer pressure high reactor trip function is Lignificantly higher for Unit I than for Unit 2.

The changes to the pressuri29r pressure high 'S' and

[

Allowable Value terms stem f rom the larger allowance for sensor drif t included in the Unit 2 Statistical Setpoint Study.

1 g+w w-p' yym*y-Mer y-eye-wy p g-

-y

&wn r y Te y9-gy.-rp-mer yo pv7,mry-myt w yw ev m-

-erm vy mm P-t-me werewr-tup-rtew p h dy-e mra e wpMw ws gyw1vwp* r Metary w-neW 4P 2m*- gr mesmg FW 4-T-tr 9'rw'

TxX-92001 ATTACHMENT 3

PAGE J Or 8 S ECT 10id : 2.2 and BASES 2.2 GM PEG U JUSTlflCATION 3-C 20 AC Unit I uses Barton pressure.ransmitters.

kosemSunt pressure transmittets are used in Unit 2.

Changes in the ambient temperature and pressure have slightly greater effects on the Rosemount transatitters than on the Barton transmitters Therefore, the *2' term will be slightly larger fer Unit 2 than for Unit 1.

The dif f erence in t he

'S' and Allowable Value terms. lies in the slightly larger allowance for the sensor drift included in the Unit 2 Ltatistical Setpoint Study.

Mit I uses the SIDP.

In this methodology, a minimum RS loop fl ow (thermal design flow) is used which is si gn i f i c a n1 l y lower than the minimum best estimate RCS loep fits ena more than bounds the measurement uncertainties.

For Unit 2. the ITDP is used.

In this methodtlogy. the minimum measured RCS loop flow is used.

The minimum measured RCS loop flow is derived from the minivum best-estimate RCS loop flow. The allowances f or the system parameter sensitivities are included in the calculation of the design limit for the WRB-1 correlation, when used with the ITDP.

The minimum measured loop flow is greater in magnitude than the thermal design flow.

3D 25 A2 The two units at CPSES have different model Steam generators, with different internal geometries.

The design dif f erences equate to a need f or a unit dif f erence to exist in the reactor trip setpoints associated with steam generator water level.

3E

? ?+4 The WRB 1 DNB correlation and the ITDP with a DNBR saf ety 4i unalysis limit value cf 1.49 15 used for Unit 2.

The W 1 DNB correlation anc the STDP with a limit value of

.c; 1.30 is used f or U.11t 1.

The phrase *the DNBR safety analysis limit' will be used to replace the rumeric limit value 50 the BASES will apply equally to both unitt.

WM

%$7 n.o TABLE 2.2-1 n

"' 53 REAC10R TRIP SYSTEM INSTRtNENTATION TRIP SETPOINTS "7

omo g

ce g

TCIAL SENSOR u,

Att OldMCE ERROR g

FLWCiIONAL UNIT (TA)

I (5) iir;p g;porf-J 34Et41E VAtUE n

1.

Manual Reactor Trip N.A.

N. A.

M. A.

W.A u

t c 2

2.

Power Range, Neutron Flux

-4 t.

~

a.

High Setpoint 7.5 4.56 1.25

<109% of RTP*

111.7% of RTP*

b.

Low Setpoint 8.3 4.56 1.25

<25% of RIP *

<27. 7% of RIP

7 3.

Power Range, Neutron Flux, 1.6 0.5 0

<5% of RIP

with 5.3% of RIP

with 6

High Positive Rate a time constant a time constant

>2 seconds

>2 seconds y7 4.

Power Range, Neutron Flux,

1. 6 0.5 0

15% of RTP* with

<6.3% of RIP

with High Negative Rate a time constant a time constant a

>2 seconds

>2 seconds 5.

Inter 1 mediate Range, 17.0 8.41 0

125% of RTP*

$31.5% of RTP*

Neutron Flus 6.

Source Range, Neutron Flux 17.0 10.01 0

$105 cps

$1.4 x 105 cps 7.

Overt rature N-16 5_a 1 65 1 3+0_a W w mate I seeNote_2]

3-A

O 476

' U * # 3 '* * "

' " N" *

I 5 # " #M#Z g'4'. 0 8.

Overpower N-16 1.93 0

$ 12% of RTP"

$115.1% of RIP"l 1

$, $:,'?k m.c Lcs

t. c r e esh) 4nZ*K e4 GTP' r*4. H J CTP' 9.

Pressurizer Pressure-Low J4 0.71 2.0

?l880 psig

?I863.6 psig l

- u,.a 1

Q.4 3-B

b. a.a z

4. 4 e.<2 z.e

?rsse m;c b M t - cm,

s 2

D.

Pressu a.u,...rizer Pressure-tiigh

7. 5 5.01 1.0

$2385 psig 12400.8 ps TJ l

o L

T L f L ; e 2,

RIP = RATED THERMAL POWER M

D'}

g (1) 1.2% span for delta-T (EtDs) and 0.8% for pressurizer pressure.

W 1.c % syn for N -11: pn

...e.,., 4 er ;

L38 %.

(c -- T..s t RTDs aJ

0. 9 & %

4cp,,..,,,,,,,,,,,,,,,,,,,,,.

av

(.3) :. c 9a sp f.- N-l(c fu.

c m e,. 4. r.. n al G O S ',', for K.gj RTDs.

e

p

-(

I 3,30 I

c - x.

m>

n,o IA8tE 2.2-1 (Continued)

" $ 's 1

J dEACTOR TRIP SYSTEN INSTRUENTATION TRIP SETPOINTS

'5~

i o o ca 1

!'g TOTAL SENSOR t.,

ALLOWANCE era 0R

,,]g FUNCTIONAL UNIT (TA)

Z, (5)

TRIP SETPOINT All0WABLE VALUE i

js 11.

Pressurizer idater Lowel-Nigh. ' 8.0 2.18 2.0 1921 of instrument 193.'3% of instruent !

e q

l

<t. da f

<-=

g 5 3~0 j

sn2n

h. u :12.

F. D 2.35 2.D s u t & wn. ~r

=.w,,

2.

Reacto. Coolant. Flanr-Law

2. 5 1.18 0.6

>90% of loop

>88.6%;o,f5 g, __ f y,.

m ;,-

~

loop S. d a' + 1

._ 3es,igrtiIow"*_._ _ aesign flow"*

i 3',

6. it,,;, z

'l. 5

t. Z S 0.87 l

3.

Steam Generater Water 25.0 22.08 2.0

>25.0E of narrow

>23.1% of narrow 1

Level - Lont-Lont g

[

range instrument

~ range instrument g

a. da:t I snan

[

-y."__

g,,

a, 4 g 3 S.*l 12.L 20 14.

Undervoltage - Reactor

[7.7]

[0]

>p30] volts-GI5dvolts-i I

i-Coolant Pumps each bus each bus

[

15.

Underfrequency - Reactor-f.4]

(0)

[0]

k7.2]Hz

>f7.1)Hz Coolant Pumps t

i 16.

Turbine Trip 1

3,_59]ps ig d6.6}psig f

a.

Low Trip System Pressure N.A.

N.A.

b.

Turbine Stop Valve M.A.

M.A.

N.A.

[lhopen hopen

~

Closure p

17.

Safety Injection Impet N.A.

N.A.

M.A.

N.A.

M.A.

from ESF 2 35.1 % et.w.m.

? J L -f % e 4 - ~

k

,,,r,..~*

J

- Lcop Sesign flow = 95,700 gpe.

r

.j,-

m. p z

s E

c--

4..-

g L a y, m.o. n m e. g., r u l ( tw

' n,5&o y,n

=

210 %

^I '" P

- f 7 p. e

,4 lu g a..,..,-

-e*5++8 1,

,,,,,,,..s.<.I F%)

tw" o.

e I

a

--)

TXX. 92001 ATTACH!ittli 3 PAGE 5 0F 8 r

u 6

0.

o e.

b I,

o o

w a

u a

4 e

- f 5

0

~

s k

)

~

e s

a

~a

~

[

E

'I 1

e

+

}

C w

y 5

3

-5 "c.

.5

.5 m "N.

s.,-prssf. -

v g

w 3,/*94

, N h,,g e

D v1.*

's I

v M~ y ',i N

8 t-e.

t i 'I "9 dB 5

g J,h N

==

edi

.f.l

$ a',tw *i I. q u s g C]1'd v o

=

g

.tk~

e.,i.-.n -- o A 45 fi JQ

- a\\ m..n.

=wt

/r 1

E.i 4 e0 '. s a*A g.5 u, g E.,

% e.

u 3

a.

x c.

a.

w, s

~.s i

s

~

"3 at e

s S

.,4

%s %

y C

4

% e w5 o.I Ee l

o 4,

I

=

s e m e +o aw

=

i o

.;\\:

w

+

n o.

.g..5' I.s

-o a

5-ri E

Q e h== ** O O m ~ u u \\ c. - o.

s-e e

4

_ f

.s n.

n t

u u

?

n a

ee m

n n

en en se P 8-5-

+

uv n

m w

w -

w e

a'

.y.

G e

n

,,t

-f.

W w

ao.o

-q i

4

.E~

C0Y.C: 3 5 ?i.;~. - ;':'T

. ' ~

1 TxX.92001 ATTACHMEf4T 3 PAGE 6 0F 8 5

s n

7 2

~

=

~

x r..

D.

.: e'

.D O

O

)

u

~y

3

= 6

= 5

)

I*

r I

s s

e 2

s

.z

.s

.K a

I :

.- e

,e

.=

t

=

.L

- =.

et

[.

y 2

v=f r

e a

3 4 (J

& g I r.-

t :

1**.*.

~. ",,

./. *

E"

=

r f

4 3 L k

w A

I 2

"E

- 4 d.e

1.. "

T;

.O

.r

.I

=..

s e w

.z

~

x~

x

?' N

{

l 4

'I et t *:

j 3

r8 E

.E.

.f, x

g

t g

4

.f.

-3

(.

4

.

T..

=

-l

.r E

E'%

t

2 :

w, g

T.

= =**

O.

1. *=

.u l.. * =

i 1 s

6 s

a d

mr Q 9

2: -

i ub

= =E t

ci

.a u

o k

g a

-4

~ g g

g ;c

1 4

a

. =,

=

E' 1

g h

1 1

C 4

=

3

3.

1 2.*:

2 +

O'

?

=.

2..

i w

' o r

2 y

v k.

3

=

a

.JE Dg k

k **

3 i

.s

=

L a.

YL

.5

~

g "a

I I

"l r

r= f1 z-z -

- 4

't 2 f **

1.

s

== -

.x..

[ q,

~

g.

~

t "I

k i

..k. =q t

=

1"

.".,* 0" t [X k j Qll m

j L

i

=.

e, sa.

e.

s

== #

< f

L

- 9 g

,^ f 0

Oy aa.* 1

.E g

i

=

.+ r

-r +

i

}C w:

h

' T'

\\

a y 5 9.T%

r u

m f w.>

r Le 1 i $)g.

st f

2.

C.

A 4w O

Q w

w f

d hb 7

r w

2 L..

2 4

2 0

\\/\\

".- J.

t*w t.

6

=

'l"

='

N

+

CH N

E k

i

...b>.

.C A

4.

L e.

w 1

_ _ =

TXX.92001 ATTACH!'ENT 3 PAGE 7 Cr 8 M

[3

.D.

1.

.C.

.O.

~

=

.,3 h m.

.=

5

. ' j 2 X C M

~

- r

~

f.

-2

.z.

O g

=

,,. =

u

. !N.

5 :

Ne n

= y, d >

v;

.P C

2

=

~

~.

3%

g 4

.e a

b h

t f (t.

V\\

N

~

=

p.

I e

b e

t i

W h

h I..

d 4

.1 p

r M.

M.

W

. =

g g

". " g"

.~

.J f

J

{

+*

1

.x 1

e Y

\\.r i.

C

,. h 3

=

~*

~~

y

- . ~

~

n n

9 5,

a a

e

.~

D a

j y

g

=

n.

A e

=

3 5

3 i

f. "i 4

.: 7 3

g

.d 3 0 W

ep Q.

=

~

=

~.

A 2 ".

." C 3

A 5

t 1

~

O

$ ". * ~

i.

-8.'

M

' s g

.W g

g d

<C 3

... -d t',.

y

."C

'J

=.

~)

m A

.m m

a.

.m

>m W

W V

i I

l

.** e.*,

+

F S

f g,'

g g

T xX-92001 ATTACHMEtti 3 PAGE O Of 8

.:u!*:N3 i Ti'+ !<5*iu !!**:NGi ti$Ct

! :*:n *::o 3 5 Eu :N3*:WrN m ::N ;! 00:N's

e n t i r,,,

)

re a" :ws eact:r tr$o cir: wits automatically c:en t*e React:r t"o i

creaners.aerever a coreition monitorea of the avactor Trio System reac es a oreset ;r calculated level.

In accition to recuncant channels arc ta cesign 4:proaca provices a Reactor Teto system.nich monitors neercus system

as, t e varisoles, therefore crovicing Trio System functional civersity,

' e < net':es' cacacility at the specifiec trip setting is reovired for those anti:iontory or civerse Reactor trips for nich no cirect crecit was assumec in t*e afety analysis to ennance the overall reliacility of the Reactor Trip System.

- e Reactor Trip System initiates a Turbire trio signal whenever Reac',or trip is initiatec.

This prevents the insertion of positive reactivity that movic otherwise result fr0m excessive Reactor Coolant System coolco.n anc thus sv:ics unnecessary actustion of the Engineerec Safety Features Actuation 5jstem.

Manual Reactor Tefo, The Deactor Trip System includes manual Reactor trip capability.

Power Darce Neutron Clua In ea:n of the Po er Range Neutron Flux channels nere are two-iacecencent Distacles, each with its own trip setting used for a

'gh and Low Range trip setting.

The Low 5etpoint trip provices protection c 'ing suberitical are 10 oc.er operations to mitigate the consequences of a po.er excursion oeginning fecm low co er, and the Mign Setpoint trip provides protection curing oo.er coerations to mitigate the consequances of a reactivity excursion from all po.er levels.

The Low 5etpoint trip may be manually blocked above P-10 (a cower level of approximately 10% of RATED THERMAL POWER) anc is automatically re estatec celow the P 10 Setp int.

Power Range, Neutron Clux, High Rates The Power Range Positive Rate trip provides protection against rapic flut increases which are C0aracteristic of a rupture of a control rod drive rousing.

Specifically, this trip complements the Power Range Neutron Flux High and Lcw trips to ensure that the criteria are met for rod ejection from mic co er, The Power Range Negative Rate trip provides protection for control roc crop accidents.

At high power a-single or multiple rod drop accicent C0uld cause local flux peaking which could cause an unconservative local DNBR to exist.

The Power Range Negative Rate trip.ill prevent this from occurring by tricoing the

reactor, No creoit is taken for operation of the Power Range Negative Gate tri; for those control roc crop accicents for.nich CNBRs wii! te greater inEn L Dr

~

21 CCMANCHE EAK - UNIT 1

-B 2-4 O R D ',* D _

g,,g ans tp os h en:1

]

i i-h TXX 92001 i

ATTACHMENT 4 PAGE 1 0F 5 4,

a i

SECTION, 3/4.0 and BASES 3/4.0 l

11G PAGE #

211ST IFl( A_1103 j

4*A 3/4 0 1 (Byron pn ;edence) The dual unit Lifniting Condition f or j

B 3/4 0 3 Opcration and the associated BASC$ are added to provide j

app'4cability definition for the Limiting Condition $ for Operation.

1 1

5 Added a discussion of valve identification for dual unit Technical $pecifications to BASES.

l 1

l I

i l

I 4

4 r

1

}..

I I

I i

t t

i i

t d'

i --

i i --

I

=-

TXX.92001 ATTACHMENT 6 PAGE 2 0F 5 2A.:u!*:'.3 M$t:*::45 *CR t* ERA?!CN A D SURVE!L'.ANCE 3E0V!REMEN'$'

3/A.S Apot:CABILITt 1.:u!':NG ::NCITION FOR ODERATICN L 0.1 00-oH nce.ith the Limiting Concitdons for Operation contained in tae i

sw::eesin; i:e:4'i:ations is reouired curing the OPERATIONAL MODES or otrer encstions s:ecifdeo tnerein; eacept that upon failure to meet the L' tit'ag j

Conoitions der Operation, tne associated ACilCN recuirements snall te met.

I 3.0.2 Noncoroliance with a specification shall exist when the recuirements :f tne Limiting Concition for Operation and associated ACTION requirements are not met within the specified time intervals.

If the Limiting Condition for Operation is restoreo crior to empiration of the specified time intervals,

ompletion of the ACTION requirementf is not recuired.

3.0.3 When a-Limiting Condition for Operation is not met, except as proviced in tPe anscciatec ACTION requirements, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months action snall be initiated to clace tne unit in a MODE in which the specification does act apply by olacing it, as applicable, in:

At least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , a,

At least-*0T SHUTDOWN.ithin the following 6 hourt.

o.

ano At least CCLD SHUTOCWN within the subsequent 24 nours.

arere corrective measures are completed that permit o:+ ration under the ACU N

requirements, tne action may ee then in accordance w n the specifiec time

. limits as measures from tne time of failure to meet t:

Limiting Condition for Coeration.

Exceptions-to these requirements are state. in the incivicual-s;e:4fications.

Tnis specif':stion 's not applicaele in MODE 5 or 6.

3,0.4 Entry into 2n OPERATIONAL MODE or other specifiec conoition snal' n:t-

[

ce maoe anen the conditions for the Limiting Conditions for Operation are n:t met ano'tre associated ACTION requires a snutdown if they are not met aitnin a specifito cime interval.

Entry into-an OPERATIONAL MOO! or specified-concition may co made in accordance with ACTION requirements wnen conforman:e to them permits continued operation of the facility for an unlimited peri:c-of time.

This provision shall not prevent passage througn or to OPERATIONAL M00E5 as-recuired to comply with ACTION requirements.

Exceptions to tnete reagirements are stated in the individual specifications.-

4A f1*n.s e r e A]-^

ua;rg: ga, q q]* ;

3/4 0 1

.. L -. -.. - - - --- - - - - - -

TXX-92001 Page 3 of 5 Insert A 3.0,5 Limiting Conditions for Operation including tne associated ACTION reauirements shall apply to each unit individually unless otrerwise indicated as follows:

Wnenever the Limiting Conditions for Operation refers to systems a.

or components which are shared by both units, the ACTION requirements will apply to both units simultaneously, unless specifically noted otherwise, and will be denoted in the ACTION section of the specification; b.

Whenever the Limiting Conditions for Oceration applies to or:y one unit, this will be identified in the A/FLICABILITY section of the specification; and

(

Whenever certain portions of a speci'ication contain operating c.

parameters, setpoints, etc., whic" are different for each unit, this will be identified in parent.neses, footnotes or body of the requirement.

d

TU-02001 ATTACHMENT 4 0Fg,L7Ag!, ;g BASES Tmerefore, if remedial measures are completed that wo t mit a return to 20WER ::eration, a penalty is not int ;rred by having to "<acn a lower MCDE of peration in less than the total time allowed.

Th'e same principle applies with regard to the allowable outage time limits of the ACTION recuirements, if compliance with the ACTION requirements for one specification results in entry into a MODE nr condition of operation for another specification in which the requirements of the Limiting Condition for Operation are net met.

I If the new specification becomes applicable in less time than specified, the difference may be added to the allowable outage time limits of the second specification.

However, the allowable outage time limits of ACTION requirements for a higher MODE of operation may not be used to extend the allowable outage time that is applicable when a Limiting Condition for Operation is not met in a lower MODE of operation.

The shutdown recuirements of Specification 3.0.3 do not apply in MODES 5 and 6, because the ACTION reouirements of individual specificati5ns define the remecial measures to be taken.

Specification 3.0.4 establishes limitations on MODE changes when a Limiting Cono1 tion for Operation is not met.

It precludes placing the facility in a nigher MODE of operation wnen the requiremats for a Limiting Condition for Operation are not met and continued noncompliance to t*ese conditions would result in a shutdown to comply with the ACTION require ents if a change in M00E5 were permitted.

The purocae of this specificat":n is to ensure that facility operation is not initiated or that higher MO:E5 of operation are not entered wnen corrective action is being taken to obta, compliance with a speci-fication by restoring equipment to OPERABLE status or parameters to specified limits.

Compliance with ACTION requirements that permit continued operation of the f acility for an unlimited period of tim provides an acceptable level cf safety for continued operation without regard to the status of the plant oefore er after a MODE enange.

Therefore, in this case, entry into an OPERATIONAL MODE or other specified condition may be made in accordance with the provisions of the ACTION requirements.

The provisions of this specifi ation should not, however, be interpreted as endorsing the failure to exercise good practice in restoring systems or components to OPERABLE status before plant startup.

When a shutdown is required to comply with ACTION "equirements, the provisions of Specification 3.0.4 do not apply because they would delay placing the facil-ity in a lower MODE of operation.

Specifications 4.0.1 through 4.0.6 establish the general requirements acplicable to surveiliance Requirements.

Inese requirements are oased on the Surveillance Requirements stated in the Code of Federal Regulations,10 CFR 50.36(c)(3):

3 4 A)

In.s e ct 8]

COMANCHE PEAK : UNIT 1 B 3/4 0-3 I

1 o

- ~ ~ -.. ~..

'TXX-92001.

ATTACHMENT 4' PAGE 5 0F 5 TXX-92001

. Attachment 4 Page 5 of 5 101fI1_A 1

Srecification 3.0.5 delineates the applicability of each specificatial to-Unit 1 and Unit 2 operation.

T The valve identification numbers (tag numbers) contain a unit. designator as

.the first-character, i.e, 105-8455 would be a Unit 1 valve with 2CS-8455 being the corresponding Unit 2 valve, The dual unit Technical Specifications i

utilize a convention of identifying valves; without_the unit-designator if the remainder of the-tag number is applicable to both units, with the unit designator if the tag is only applicable to one unit.

a When a specification is. shared per 3.0.5a. the ACTION section contains the id9ntifier "(Units 1 and 2)".

I I

4 6

l

TXX 92001-ATTACHMENT 5 PAGE 1 0F 5 SECTION: 3/4.1.2 111M PAGE #

JUSTlFICATION 5A 3/4 1 11 The CPSES boric acid system is composed of 4 boric acid 3/4 1 12 A5-transfer pumps (2 dedicated to each unit). 2 boric acid storage tanks and associated suttion and discharge piping to allow any of the four pumps to be aligned to operate with a borated water supply of E.ther tank.

)

The-requirement of the specification is to maintain an adequate minimum borated water volume.

CPSES desires

- to meet the-minimum volume requirement f or both units from either the volume of-a single boric acid storage tank or from the combined volume of both boric acid storage tanks.

This plant specific configuration is acceptable because boric acid flow paths are required to

-be OPERABLE in accordance with Technical Specifications 3/4.1.2.1 and 3/4.1.2.2.

The change is deemed appropriate to prevent unnecessary-plant shutdowns.

CPSES could encounter an-operational condition (i.eL one inoperable tank), that would require a unit shutdown even though an adequate borated water

! supply existed for both units in the other OPERABLE tank.

The footnotes have been added to envelop the anticipated operational situations while maintaining the existing specification format and content.

Because the LC0 relies on both shared equipment (the boric acid storage tanks) and unit specific equipment 4

(e.g. pumps and Refueling Water Storage Tanks), the requirements of-proposed specification 3.0.5a.-are not-i-

applicable.

N j

i

.m

TXX-92001 ATTACHMENT 5 PAGE 2 0F 5:ga; ;7;, ::y ::.

3 gug 3CRA'E0 aAIER 30L'ECE - i=UI*C'aN

,!Ml?!NG CON 0I'!!N ::E l0E:aTION 3.'.2.5 as n Ti-4

.?

.,e o f : e 'cl' ing :: rated -ater s:ur:es sna t '

te 2EE53i3:

a.

A Oo-ic 3: 0 3 Orage tant wit 9:

4

'. )

A T *

i ' L ?. irdicated 00ratec.ater level of 10%s,ren wSing t e

or': a:tc transfer cumo, toratec.ater'evalof20*G7)

)

Atinimum incicate:

tween. sing e

grav1;y feeo connection, 3)

A minimum boren concentration of 7000 ppm, and A)

A T,i,' mum solution temperature of 65*F.

Tre ef;ei' g ater storage tans (RWST) with:

1)

A minimum indicated corated water level of[2h;,

2)

A T'.nimum ocron concentration of 2000 ccm, ano 3) a mi-imum solution temperature of A0 :

A??LICAEILIrt:

"C:E5 i and 6.

-u..N:

o.,

a'itn no baratec.atea source OPERABLE, susoend all operatiens involving CORE ALIDAT g r h tivi t i O

Sa-e;+, g

.w.

- - ~ __ __

.s applo'c b fe_,,

aee not 4.1.2.5 The above recuired corated ater source shall te demonstrate:.PERASLE:

At least once per 7 days t/:

a.

1)

Verifying the boren concentration of tne wate,

2)

'/erify rg the indicated berated water level, and i

3)

Verjirg the boric acid storage tank solution temperature nen it is t~e source of borated water.

t b.

At least once per 2A hours by verifying the RWST temoerature when it is tne 30erce of boratec water and the outsice air temperature is less than 20*.

.lsset k CCMAN HE EAK - UNIT 1 3/4 1-11 I

TXX-92001 ATTACHMENT 5 PAGE 3 0F 5 T. NS E RT

With both units in MODES 5 or 6, one boric acid storage tank with a minimum indicated level of [13]% or two boric acid storage tanks totaling a minimum indicated level of [20)%, ensures a sufficient volume for both units.

With both units in MODES 5 or 6, one boric acid storage tank with a minimum indicated level of [32)% or two boric acid storage tanks totaling a minimum indicated level of (40]%, ensures a sufficient volume for both units.

1

TXX-92001 ATTACHMENT 5 PAGE 4 0F 5 REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.6 as' required by Specification 3.1.2.2:As a minimum, the following borated water a.

A boric acid storage tank with:

bIl y

+-

1)

A minimum indicated borated water level of 50 b'

+

/

2)

A minimum boron concentration of 7000 ppa, and

\\

3)

A minimum solution temperature of 65'F.

I b.

The refueling water storage tank (RWST) with:

1)

Amin 1mumindicatedboratedwaterlevelof[9h, 2)

A boron concentration between 2000 ppe and 2200 p m,

3)

A minimum solution temperature of 40*F and 4)

A maximum solution temperature of 120*F.

APPLICABILITY:

H0 DES 1, 2, 3, and 4.

ACTION:

\\

With the boric acid storage tank inoperable and bein3 used as one a.

{

of the above required borated water sources, restore the tank to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY with i

the next 6 Lours and borated to a SHUTDOWN MARGIN equivalent to OPERABLE [ status within the next 7 days or be in CO atleast1.3)Ak/kat200*F;restoretheboricacidstoragetankto 1

the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

i

(

b.

With the RWST inoperable, restore the tank to OPERABLE status i

within 1 hours1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

pmveper/ov v

erm,,

n_,

\\

k presta ca.s e f Speci fi e cA tau 3. 0. h

~ r e moi app Ite ed j e,

({Lsear L) 4

%x

~

mm-COMANCHE PEAK - UNIT 1 3/4 1-12 Amendment No. 5

ATTACHMENT 5 PAGE 5 0F 5-INSERT L

With both units in MODES 1, 2, 3 or 4, one boric acid storage tank with a minimum indicated level of [87]% or two boric acid storage -

_ tanks-totaling a minimum indicated level of [100]%, ensures a sufficient volume for both units.

1. - With one unit in MODES 1,2,3 or 4 and the other unit in MODES 5 or-6,- one boric acid storage tank with a minimum indicated level of

[62]% or two boric acid storage tanks totaling a minimum indicated level of [70]%, ensures a sufficient volume for both units.

?

=

1

TXX 92001 ATTACHMENT 6 PAGE 1 0F 3 SECTION:

BASES 3/4.2.

3/4.2.2 and 3/4.2.3 ITEM PAGE #

J.IIST I r 1 C AT 104 6A B 3/4 2-1 A6 The WRB-1 DNB correlation and the ITDP with a DNBR safety analysis limit value of 1.49 is used f or Unit 2.

The W 3 DNB correlation and the STDP with a limit value of

1. 30 i s u s ed f o r 'Jn i t 1.

The phrase "the limit value' will be used to replace the numeric limit value so the BASES will apply equally to both units.

6-B B 3/4 2 4 Al The DNBR margin values which are given f or U:iit 1 are not applicable to Unit 2.

The appropriate BASES f or Unit 2 values will be provided later.

TXX-92001 ATTACHMENT 6 i

PAGE 2 0F 3-3/4.2 POWER DISTRIBUTION LIMITS BASES m

+ke bW value The spepffications of this section provide assurance of fuel integrity during Con tion I (Normal Operation) and II (Incidents of Moderate Frequency) events b.

(1) maintaining the minimum DNBR in the core greater than or equal the fission gas releaseuring normal operation and in short-ters transients, and (2) to properties to within ass,umed design criteria. fuel pellet temperature, and claddi In addition, limiting the peak linear powvr density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200*F is not exceeded.

The definitions of certain hot channel and peaking factors as used in these specifications are as follows:

F (Z)

Heat Flux Hot Channel Factor, is defined as the maximum local heat 9

flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for sanufacturing tolerances on fuel pellets and rods; and Fh Nuclear Enth'alpy Rise Hot Channel Factor, is defined as the ratio of the integral of linear power along the rod with the highest integrated power to the average rod power, 3/4.2.1 AXIAL FLUX DIFFERENCE The limits on AXIAL FLUX DIFFERENCE (AFD) assure that the F (Z) upper bound 9

envelope of the F limit specified in the CORE OPERATING LIMITS REPORT (COLR) q times the normalized axial peaking factor is not exceeded during either normal operation or in the event of xenon redistribution following power changes.

Target flux difference is determined at equilibrium xenon conditions.

The rods may be positioned within the core in accordance with their respective insertion limits and should be inserted near their normal position for steady-state operation at high power levels.

The value of the target flux difference obtained under these conditions divided by the fraction of RATED THERMAL POWER is the target flux difference at RATED THERMAL POWER for the associated core burnup conditions. Target flux differences for other THERMAL POWER levels are obtained by multiplying the RATED THERMAL POWER value by the appropriate fractional THERMAL POWER 1evel.

The periodic updating of the target flux difference value is necessary to reflect core burnup considerations.

COMANCHE PEAK - UNIT 1 B 3/4 2-1 Amendment No. C

1 IXX-3hC1 ATTACHMENT 6 PAGE 3 0F 3 power DISTR!suTION LIMIT'S BA)ES HEAT FLUX HOT CHANNEL FACTOR and NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued) and minimum DNBR are not exceeded and (2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200'F ECCS acceptance criteria limit.

Each of these is

-murab'a but will normally only be determined periodically as speM' in Specifications 4.2.2 and 4.2.3.

This pariedic surveillance is sv. -

' asure *F t the limits are maintained provided:

4 a.

Control. S

'1* t oup move together with no individual red insertion e than t 12 steps, indicated, from the group dementi b.

Control rod gm -

> u?r witr overlapping groups as described in Specification /

c.

The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained; and d.

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

FhwillbemaintainedwithinitslimitsprovidedConditionsa.through TherelaxationofFhasafunct'enofTHERMALPOWER

d. above are maintained.

allows changes in the radial power shape for all permissible rod insertion limits.

Fuel rod bowing reduces the valL's of DNS ratio.

Credit is available to

,pf(set this reduction in the generic margin.

The generic margins, totaling L9.1)E ONBR completely offset any rod bow penalties.

This margin includes the following:

a.

Design ifnit DftBR of 1.30 vs 1.28, b.

Grid g ing (K,) of 0.046 vs 0.059, f

T y,. Diffusion Coefficient of 0.038 vs 0.051, 6$

c.

d.

D wit Multiplier of 0.86 vs 0.88, and a.

Pitch reduction.

J The applicable values of rod bow penalties are referenced in the FSAR.

COMANCHE PEAK - UNIT 1 8 3/4 2-4 Amendment No.

1

TXX 92001 ATTACHMENT 7 PAGE 1 0F 3 SECTION: ~3/4.2.5 and BASES 3/4.2.5 1T13 f_ AGE #

JUSTI F IC AT IOJ 7-A 3/4 2-12 The WRB-1 DNB correlation and the Westinghouse-improved

.B 3/4 2 6 Al Thermal. Design Procedure (ITDP) is used in Unit 2, while the W-3 DNB correlation : and the Westinghouse Standard Thermal Design Procedure (STDP) is used in Unit 1 in design of the core, The different DNB correlations and core dasign dif ferences result in the allowable values-of the DNB relued parameters _,'eing dif ferent.

7B B 3/4 2-6 Al The WRG-1 DNB correlation and the ITDP with a DNBR safety analysis limit value of 1,49 is used in Unit 2. L The'W DNB correlation and the STDP with a limit value ef-1.30 is used-in Unit-1, The phrase "at or above the-limit value" will be used to replace the numeric limit value so the BASES will apply equally to both units.

L -

-w.

,,v

~e v

1

-TXX-92001 ATTACHMENT 7' POWER DII'4!!UT!CN LIMITS

~

PAGE 2 0F 3 3/4.2.5 ONB PARANETERS LIMITING CON 0! TION FOR OPERATION 3.2.5 The following DNBarelateo parameters shpil-be maintained witnin the statec limits:

Incicatec Reactor Coolant System i""91 kwfff f a.

< 592'F for unit 1

-~3 7-A Indicated Pressurizer Pressure 12207 ps[igarer un,e i ee unit 2 t O.teil pu g

Re unit-2 Indicated Reacter Coolant System (RCS) Flow 1 389,700 gpm " for und 1 c.

a APPLICABILITY:

MODE 1.

3Y' ACTION:

~~~

With any of the above parameters exceeding its limit, restore the parameter to within its limit witnin 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE RE0VIR2MENTS 4.2.5.1 Eacn of at ' east once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .tne above parameters shall be verifies to be witnin its limits 4.2.5.2 The RCS total flow rate snall be verified to e within its limits at least once cer 31 cays by plant comcuter indication or measurement of the RCS eloow tan sifferential pressure transmitters' output voltage.

4. 2. 5. 3 Tne RCS loop flow rate incicators shall be subjected to a CHANNEL CALIBRATION at least once per 18 montns.

The channels snall be normalize based on the RCS flow rate determination of Surveillance Recuirement 4.2.5.4

4. 2. 5. 4 The RCS total flow rate shall be cetermined by precision heat calance measurement after eacn fuel loacing and prior te operation aoove 75% of RATED THERMAL POWER.

The feedwater pressure and temperature, the main steam pressure, and feecwater flow dif ferential pressure instruments shall te cali-tratec witnin 90 cays of performing the calorimetric flow measurement.

" Limit not acclicaole curing eitrer a THERMAL POWER ramo in excess of 5% of RATE: *HERMAL POWER Oer minute or a THERMAL POWER steo in excess of 10% of RATE 0 THERMAL POWER.

- !ncluces a 1.8% flow measurement uncertainty.

CCMANCwE DEAK - UNIT 1 3/4 2-12 1

TXX-92001 ATTACHMENT 7 PAGE 3 0F 3 POWER OfSTRfBUTf0N LIMITS BASES (7B af or d e M h'm it' 3/4.2.5 DNB PARAMETERS I

A VE.].

The limits on the DNB-related parame rs assure that each of the param-eters are maintained within the normal s assumedinthetransientandaccidentap/eady-stateenvelopeofoperation alyses.

The limits are consistent withtheinitialFSARassumptionsandyavebeenanalyticallydemonstrated adequate to maintain a minimum DNBR < 1. 2 throughout each analyzed

! transient.

Thejindicated T,yg -value of 592.7'F (conservatively rounded to gy 592'F) and theTindicated pressurizer pressure value of 22C7 psig correspond to analytical limits of 594.7*F and 2193 psig respectively, with allowance for measurement uncertainty.4 The indicated uncertainties assume that the reading from four channels ill be averaged before comparing with the required limit.

The 12-hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following loao changes ar.d other expected transient operation, and to detect any significant flow ci; gradation of the Reactor Coolant System (RCS).

The additional surveillance requirements associated with the RCS total flow rate are sufficient to ensure that the measurement uncertainties are limited to 1.8% as assumed in the Improved Thermal Design Procedure Report for CPSES.

Performance of a precision secondary calorimetric is required to precisely determine the RCS temperature.

The transit time flow meter, which uses the N-16 system signals, is then used to accurately measure the RCS

-flow, Subsequently, the RCS flow detectors (elbow tap differential pressure detectors) are normalized to this flow determination and used throughout the cycle.

\\

The unw 2 inclic.oded Lg value of E54e,]'F (cornermtively munded N

Y OMe-] F) and h Unh 2 indkuted pressurizer yressure valwe ba.fer.} pd3 correspewd to ana,lyticu.l Lmih of JN. 'F and (2188 psig respeciiuely, En cdicu)ance Er measurement uncertainty.

COMANCHE PEAK - UNIT 1 B 3/4 2-6 Amendment No. 1

TXX-92001 ATTACHMENT 8 PAGEJ1 0F 2 i

SECTION: 3/4.3.1

.11EM - PAGE #

JUSTIFICATION 8-A 3/4 3 6 The-valve-identification numbers (tag numbers) contain a unit designator as the first character, i'.e.1CS-8455 would be a Unit-I valve with 2C5-8455 being the corresponding Unit 2 valve.

The dual unit Technical Specifications will utilize a convention of identifying valves; without the unit designator if the remainder of'

' the tag number is applicable to both-units, with the unit designator if the tag is only applicable to one unit.

(See revised BASES for 3.0.5).

?ABLE 3. 3-1 (Continuec)

ATTACHMENT 8 PAGE 2 0F 2 4c7 ton-stA;EMENT5 (Continued)

ACTION 3 - With the numeer of enannels OPERABLE one less Channels OPERABLE requirement and with the THERNAL POWER level:

a.

Below the P-6 (Intermediate Range Neutron Flux Interlock)

Setpoint, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P-6 Setpoint, t.

Above the P-6 (Interrneciate Range Neutron Flux Interlock)

Setpoint but celow 10% of RATED THERMAL POWER, restore tne inoperable channel to OPERABLE status prior to increasing THERMAL POWER aeove 10% of RATED THERMAL POWER, ACTION 4 - With the number of OPERABLE channels one less tha Channels OPERABLE requirement, suspend all operations involving positive reactivity changes.

ACTICN 5 - with the numeer of OPERABLE channels one less than Cnannels OPERABLE reovirement, restore the inoperable enannel to OPERABLE status witnin 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or within the next nour c e tne reactor trip ertakers, suspend all operations involving positive reactivity changes and verify either valve @5-B455

'A' or -valves $5 B560, FCV-111B, @S-8439, $5-8441, ano @S-8453 are closed and securea in position, and verify this position at least once per 14 days trereafter, Wi no channels OPERABLE complete-all tne above actions within - nours and verify the positions of tne acove valves at least :nce per 14 days tnereafter.

ACT!ON 6 - With tne number of OPERABLE channels o t less than the Total NumDer of Channels, STARTLP and/or POWER OPERATION may proceec proviced the following conditions are satisfied:

The inoperable channt1 is placed in the trippea condition a.

within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and The Minimum Channels OPERABLE requirement is met:

D, o-ever, tne inoperable channe) may be bypassea for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of other channels per Specification 4. 3.1.1 ACTION 7 - With less tr.ea tne Minimum Number of Channels OPERABL 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months determine by ceservation of the associated permissive annunciator window (s) that the interlock is in its requitec state for the existing plant condition, or apply Specification 3.0.3.

COMANCHE EAK - UNIT 1 3/4 3-6

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

F TXX-92001 ATTACHMENT. 9 PAGE 1 0F 5 y.

SECTION: 3/4.3.2

'! TEM PAGE #

JUST!FICATION 9-A 3/4 3-25 The dif ference in the

'Z' term is attributable to the 3/4 3-27 A2 use of Rosemount_ pressure transmitters in Unit 2 versus the use of Barton pressure transmitters in Unit 1.

The uncertainty induced by high temperature environments i.s less for-the Rosemount pressure transmitters than for.

the'Barton transmitters, 9B

- 3/4 3-27 A2 The two units at CPSES have different model Steam 3/4 3 28 A2 Generators, with dif ferent internal geometries.

The design differences equate to a need for a unit dif ference to. exist in the ESFAS setpoints associated with Steam Generator water-levels.

9-C 3/4 3-28.A2 The dif ference in 'the *Z' term is attributable to. the -

use of Rosemount pressure transmitters-in Unit 2 versus

.the use of Barton pressure transmitters in Unit 1.

The uncertainty induced by normal changes in.the ambient temperature-is-slightly greater for the Rosemount

_ pressure transmitters than for the Barton transmitters.

9-D 3/4 3 The change to.the Allowable Value results from the allowance ftr the use of dynamic measurement methods for Unit 2, versus the static measurement methods used in Unit-1, i

l l.

i.

1 f

e.

e

~

3M a, -< x.

r 2 rv ~ h o ?a

} Al!!f _ L y_1 25~

v.

e A

I NGINII RI D SAII I Y l I AlilH15 At lHAlION *.Y'.Il H IN'elHHH1 NI AIIGN IRIP St IPOINi'.

1

.l.

SINSOR u

353 IDIAl iRROR 9

liff4Cll0NAl llNil Al l 0WANfl (IA) /

( *. )

l R il'

f l P O I N I AllOWAllil VAllit ie

,2, I.

Salcty inject ion (!CCS, Heat inr is ip, 3

lecilwater Isolation, Control Room Imergency Recirculation, imeselestry Diesel Generator Operation, Contain-ment Vent Isolation Station Servite l

Waler, Phase A isolation. Auxiliary I reilwat er-Motor Driven Pump, f ue leine leip, Component Couling Water, 1.sential Ventilation Systems, anil Cont a inment

pray 1%mep).

w s.

a.

Manual luitiatinn N.A.

H. A.

N A.

N.A N A.

y li.

Automatic Attualion ioqie N A.

N.A.

H.A.

NA te A

[

anal Attuation Relays t.

Containment Pressure--liiyh 1

/. /

0. Il

1. /

3.2 psig

1. H p. iy

_ lH/u psig Ints) s. les syl al.

Picssian i/cr Picssure--l ow gji'.se 10.91 7.0 f, l, h_

(d.cI lis. 3]

(2,01 z D LO y

ytef 3,>.fgj '

~

5(e]am I ne P essin e--l ow G{l/.l' 15.11 f

' 7. O '

T.O s's I p.o 9/

e.

y e. p.m.t C 405 p.f c578.*tp;f CuntaiYia04ii Spray

1 3 4 !5 10 7.

a Hanuaf initiation N A.

NA N A.

N A..

f4 A Anlumalis At i n.il son i ny ii.

N A.

NA NA anil Asin.stion Relays a

fs.nl a e nment P essus e -lleyli-1

?/

o. /1 I /

IH / ps os it! ti e. os

-4 R$?

1ABLE 3.3-3 (Continued) mgg n

oM8 l

ENGINEERED SAFEIY f EATURES ACIUATION SYSTEM INSTRLMENTATION TRIP SEIPOINIS-

5 ~

n 1

m SENSOR

=

101AL ERROR h

' FUNCTIONAL UNIT ALLOWANCE (TA) Z (S)

TRIP SEIPOINI All0WABLE VAIDE l

[

4.

Steam Line Isolathn x

U a.

Manual Initiasion M.A.

N.A.

N. A.~

N.A.

e b.

Automatic. Actuation Logic M.A.

N.A.

M.A.

and Actuation Relays c.

Containment Pressure--High-2 2.7 0.71,

1. 7

$.2 psig 16.8 psig 6

kh

( n 1.3 cs77.'tcE f]

d.

Steam Line Pressure--Low 17.3 15.01

2. 0

>605 psig*

>$93.5 psig*

Steam,'Line Pressure -

k

9. t 'i LD

%os ess a

8.0 0.5 0

$100'psfa^

$1/8.Tpsiaal

<p e.

4 Neoative Rate--High a D o.c 1 y

5.

TurbiNrYandFeedwater 3U 0 #5 0

^#

2 w' sia -

$ # 7sa p Isolation a.

Automatic Actuation Logic M.A.

N.A.

N. A.

and Actuation Relays 8

b.

. Steam Generator Water

7. 6 4.78 2.0

$82.4% of

$84.3% of narrow 7

Level -High-High

/

narrow range range instrument

1) f1 4 I instrument span.

q-6 snan

3) /f,3. * <L k

c.

Safety Injection See Item.1.

above for all Safety Injection Trip setpoints and Allowable Values.

3 5

e i

2 N

N' < s. 5 a.4 2.o s n.s

s s 6<.s2 s......,'.3

~

nam-cc. s;e a.. n..-

r.,

t-53~~~+

s. ~.

c S,v.*,

..md, m

m m_

m

_--_.m

/.

25s:'

%$Y TABLE 3.3-3 (Continued) 2. n ' O f*1 C ENGINEERED SAFETY ffATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS ]E n Si e $[N50R TOTAL ERROR h FUNCTIONAL UNIT ALLOWANCE (TA) Z (5) 1 RIP SEIPOINT ALLOWABLE VALUE 6. Auxiliary feedwater c5 a. Automatic Actuation Logic N.A. N.A. N. A. " N.A. N.A. and Actuation Relays H s. b. Steam, Generator Water r 25.0 22.08 2.0 > 25.0% of > 23.1% of narrow Level--Loktow . lj u a

1 h

~ narrow range range instrument \\ b u^' * ?- )/ instrument span. 9-B t span. l 35.'l

11. L

).D + l w} c. ' Safety-Idjection - Start See Item 1. above for all safety injection Trip Setpoints and ) Motor Driven Pumps Allowable Values. mm d. Loss-of-Of fsite Power N.A. N. A. N. 4. N.A. N.A. e. Trip of All Nain feedwater N.A. N.A. N.A. N.A. N.A. Pumps 7. Automatic Initiation of ECCS Switchover to Containment Sump a. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A. and Actuation Relays s b. RWST Level--Low-Low g

2. 5 0.71 1.25

- 40.0% of - 38~9% of span g o de { som b6/bEIdentWith See Item 1. above for all Safety Injection Trip Setpoints and k Safety Injection Allowable Values. 8. Loss of Power (6.9 kV & 480 V Safeguards System Undervoltage) z 8 6.9 kV Pre' erred Offsite N.A. N.A. N.A. - 5004 y < 5900 V a. Source Undervoltage [4900V m - 2 35.4 % 04 g 3 5.1 <h c I u r, c.. 2.5 C.fl] {!.1 S) ?{W Ch'c *S ? h *

7

' ~{'b " ' 'e' y.

l 3"l% ~ 'R W gy:p i r Tf5 g. I^l"l. L kl ICenti'!"e1} ~ 8 y f MGINi f RI D SAI[ I Y l l A.l.tlRI.S _Ar i. t!A_l l0N SYS__T[ M_ _IN_S___I_Ril_M_E NI Al _l0_N I_R. I__P

,1 I,P_U_IN_I S g

T" $1NSOR 101A1 (RROR R IUNCI10NAL UNII .AliOWANCE (IA) Z (5) IRIP 51IPOINI Al i OWA81[ vat 13(- 8. Less of Power (6.9 kV & 480 V g Saf eguards-System Undervoltage) (Continue:1) . ~y i

[5004]V

'nHhv li 6.9 6V Altternate Offsite N A.

N.A. NLA. Souro Ifndervoltage . wndy )

[2031V h 9 3*i V L.

6.9 kV Bub Undervoltage' N.A. N.A. N.A. - {l450 V-

[6054)V

.[5933]V d. 6.9 kV Degraded Voltage N.A. N.A. N.A.

h39]V

' @ 15]V e. 480 V Degraded Voltage N, A.; .N.A. N.A. [ t. 480 V iow Grid N.A. N.A. N. A

[44 /]V

@41]V UndervoItage r>o 9. Control Room toergency Recinulation a. Manual Initiation N.A. N.A. N.Ac M.A. N.A. b. Safety Injection-See Item-l. aliove for all Safety injection Trip 5etpoints ased - A I It w.it. l. V.. I nes. s 10. Ingineered.5afety Ieatuses-Actuation System Inte locks ,ressurizer Pres.nre, P-1I IN-A. N.A. N.A. P l'H10 psiy 19/5 2 jKijgj a. s VD n n,. + x N, n. 13 n e1. n. x nuo p) < oq w, pq b. Reactor 1 rip, P-4 N A. N. A. N A. N.A. NA II. Sol id Stat e Saf e:Jesas~sl, Seegese ens e N A. N A. N A_ N A. NA (5%5)

___..-,_....._.____._..---_._._..___..._._.m . _.. ~. _.. _. - _. n l', . TXX 92001. ATTACHMENT-~10 LPAGE 1 0F 2 SECTION: 3/4.3.3 LT13 [Afd_f! JttSTI F IC AT104 10-A 3/4 3 40. The Control Room Emergency Air Cleanup System is common to both units at-CPSES. The added notation is an-operational enhancement to identify that both units shall comply with the required ACTION. (See revised specification 3.0.5a) i i i 4 1-4 = j r -f i' k? 1 } 4 -- s 4 i I A I J

TXX-92001' LATTACHMENT 10-TABLE 3.3o4 (Continued) -.PAGE 2 0F 2 TABLE NOTATIONS Must satisfy Gaseous Effluent _ Dose Rate recuirements.in Part I f tne

ODCH, Ouring CORE ALTERATIONS or m:vement of irraciated fuel witnin esntsivent.

ACTICN STATEMENTS AdTICN 27 - With the numoer of OPERABLE channels _less than the Mini mum Channels OPERABLE requirement, operation may continue crovicea -tne containment ventilation-valves are maintained-closec. Tne-containment pressure relief valves may only be opened in com-pliance witn Specificatior 3,6.1,7 and the radioactive gaseous effluent monitoring instrumentation requirements in Part I of the'00CM. ACT!!N2[8-With the number of OPERABLE channels o Channels OPERABLE requirements, within 1 hour secure tne Control Room makeup air supply fan from the affected intake or i ni t.i s t e and maintain operation of the Control Room Emergency Air Cleanuo -System in emergency recirculation. ACTION 29 - With the numoer of OPERABLE enannels less than the Minimum Channels OPERABLE requirement, comply with the ACT!ON require-ments of -Specification 3.4. 5.1. linits ] a.n d 2) E CCMANCrE PEAK - UNIT l-3/4 3-40 I

i TXX-92001 ATTACHMENT 11 PAGE 1 Of 2 SECTION: BASES 3/4.4.1 iT111 PAGE # JUSTIFICATION ll-A B 3/4 4 1-The WRB-1 DNB correlation and the IDTP with a DNBR safety analysis. limit value-'of 1.49 is used for Unit-2. The i W+3 DNB correlation and the STDP with a limit.value of 1.30 is used for Unit 14 The phrase ' greater than or equal-to the limit value" will be used to replace the numeric-limit value so-the BASES will apply equally to both units, n 'l I 1

-TXX-92001 ATTACfiMENT 11 PAGE_ 2 0F 2 3'4.4 e?A VOR COOLANT Sf5 TEM Ba5ES ,3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is-designed to operate with all reactor coolant loops in coeration and maintain DNBR

.e 1.;7 curing all normal operations anc anticipatec transients.

ODES 1 and 2 witn one reactor coolant loco not in cieration this speci'ic ion requires that the plant ce in at least MOT STANCBf ithin 6 nours. g, g g 7 -In MODE 3, two 'r actor coolant locos provice sufficient heat removal e cacacility for removing core decay heat, even in the event of a bank witnerawal accident; however, a single ret 'or coolant loop provices sufficient heat removal capacity if a bank witho. awal accident can be prevented, i.e., by coening the Reactor Trip System breakors. Single failure constoerations require that two loops be OPERABLE at all times, In MODES 3, 4, and 5, the oceraoility of the recuired steam generators is based on maintaining a sufficient level to guarantee tube coverage to assure heat transfer capacility. In MODE 4, and in MODE 5 with reactor coolant loops filled, a single reactor coolant loop or RHR loop provices sufficient heat removal cacacility for removing cecay heat; but-single failure considerations require that at least two loops (either RHR or RCS) ce OPERABLE, In MODE 5 with reactor coolant loops not filled, i single RHR loco Drovides sufficient heat removal capability for removing decay

eat; but single failure considerations, and the unavailability of the steam ge erators as a neat removing component, require-that at least two RHR loocs be OPERABLE, The operation of one reactor coolant pump (RCP) or one RHR pump provides uecuate flow to ensurtmixing, prevent stratification and procuce gradual reactivity enanges curing boron concentration reductions in the Reactor Coolant System.

The reactivity change rate associated with boron reduction will, therefore, De within the capability of operator recognition and control, The restric qs on starting an RCP with one or more RCS cold legs less than or equal u, 250fF are provided to prevent RCS pressure transients, caused .by energy additions from the Secondary Coolant System, wnich could exceed the limits of 10 CFR 50 Appendix G. The RCS will be protectec against overpressure -transients and will not exceec the limits of Appendix G by restricting starting of tne RCP,J mto when the secondary water temperature of eacn steam generator is less than[50J above each of the RCS cold leg temperatures. c -3/4.4.2 SAFETY VALVES The' pressurizer Code safety valves operate to prevent the RCS from being pressurizec above its Safety Limit of 2735 psig. Each safety valve is cesigned to relieve 420,000 lbs per-hour of saturatec steam at the valve Set:cint, In the event that no safety valves are OPERABu., an operating RHR loop, connected to thw1CS, provides overpressure relief capability anc will prevent RCS overpressurization. 1 COMANCHE PEAK - UNIT 1 3 3/4 4-1

-TXX-92001 ATTACHMENT 12-PAGE 1 0F 7 SECTION:-3/4.4.8 and BASES 3/4,4.8 [TW PAGE # )J!1T I F1 C AT ION 12-A 3/4 4 24 -The RCS heatup and cooldown limitations for both units 3/4-4 25 ' have been evaluated and the-Unit I limitations are more-conservative in all aspects to Unit 2. In an ef fort to reduce unit differences within the Technical Specifications, it is desired to impose the more limiting Unit I curves onto both units. This change is to include the appropriate Unit 2 Material Property Basis information to make the figures applicable to both units. _12-B. B 3/4 4-6 The text is appropriate'for both units, therefore the B 3/4 4-13 elimination of the reference to Unit 1 is required,' 12-C -B 3/4 4 8 The Reactor Vessel Fracture Toughness Properties of B 3/4 4 8a* Unit 2 have been added to _the BASES for completeness, o

Proposed New Page

A TXX-92001-

-ATTACHMENT-12 PAGE 2'0F 7 19-A MATERIAL PR0cC1TY BASIS i CONTROLLING MATERIAL: LOWER SHELL PLATE R1108 1 ( UtJiT I) tNWR.MEb4TE SHEL.L 9 LATE R%C& Q (wit b. INITIAL RTNOT: 0 F (ONiT D toop (otJiT al 3 RT AFTER 16 EFPY: 1/4T. 35* F (OMIT 13 W F (OMIT M NDT 3 3/47, 70* F (UNIT 0 3 Qo P CutJIT h t CURVES APPLICABLE FOR HEATUP RATES UP TO 100* F/HR FOR THE SERVICE PER!00 L'P ': 16 EFPY. CONTAINS MARGIN OF 10'F AND 60 PSIG FOR POSSIBLE INSTRUMENT ERRCRS. 2500 '~ r i ~ .' UNACCEPTABLE i**** 'OPERAf!0N LEAK TEgf l '[ jl 2250 L;Mt! i i i i it i,, .i 1 i r i4 i i n .>i I I f 2000 I III [ALIIY (l.9lI '3 'l 'l BASE: ON mEATUP 7 CURvt CF UP c a l 0 60*F/*R i g w ' 3 i f I l. 1 100*F/HR I 1500 l ^, HEATUP I ' ' ' f f e RATE 5 t UP TO /' j g. r 1250 o 60*F/NR I l v i ' w f 1 E 100* F/HR ll 1000 6 7. /,' ,I ll', l o i.,. a 3 1 i < /1 I > i r' I 1 I M0 l z i .4 C2!!!CALITY LIMIT SA5ED ON 'i !NSERY!CE HYORCSTATIC TEST " TEMPERATURE (22)*') 'OA TWE E SERV!CE PERIOD UP T3 16 EFPY-t I I I f I_ I 250 l l AC:EstaeLE l OPEGAf!;M 1 e 0 50 100 150 200 25c J00 350 400 450 500 IMON:ATED TEMPER ATURE (DEG P) FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO is EFPY COMANCHE PEAK. UNIT 1 1/4 4-24

TXX-92001 ATTACHMENT 12 PAGE 3 0F 7 un'E : 2L, ne:Etiv ustsf ~% s [e IOTEEMEDt6TE SHELL PLtTTE R38Cl 'l(UMIT d) CONTRCLL!NG MAT!;.'AL: L;'4EA Sggt; ;LAI! R1103-1 (Utair 13 3 !N!TIAL RTNDT: 'O'T (uta rig iocF (uurt f) r RT AFTER'15 EF?Y: 1/47, 85'T (Ourt h 31 op (a.r e / NOT 3 3/47, 70'Ft00ir 0 3 WP (cotr $) CURVES AP8'.! CABLE :CR COCLOO'4N RATES UP 70100'F/HR FOR THE 3E;y;- 70 15 EFPY. CONTAINS MARGIN CF 10*F AND 60 PSIG FOR pCSSIBLE " ~ NS*:t

ERRCRS, 2500 -

, WAC:!P*A.1L2 f CFtufi:N

50 -

i s 4 l300 o o i ,.5g. 1 l = ~ 500 - i 3 i 4 ?. ':50 2 i e '000 ' 4 5 C00LDCME s urts '30 2 ('TAUL) Wees u - e 'auem [ 20 e 'O H/ 500 ~,A 60 "r i== 100 50 ^- i t A CC E F*.A3 L,1 o-eecu :::.s 1 10 100 ?50

go l50 300 150 40 0 450 530 NoiCatt3 fra,1staATL81IQtG 8)

FIGURE 3.J-3 DEACTOR CCOLANT S rSTF.M COOLDOWN LIMITATIONS-APPLICABLE UP TO ;5 trPY I ~.OMANCHE PEAK 'J N I T 1 3/4425 y

TXX 92C. ATTACE AT 12 PAGE 4 0F ' ga$El 3/4 4 e PRE 55URf/TEMPERAtycE LIu!?5 The tercerature and pressure charges during heatuD and Cooldown are L 15tted to ee consistent with the recuire*ents given in the A5ME Boiler and Pressurt. Vessel Code, Svetion !!I. Accendia G and 10 CFR 50 Appendia 0. 1, T*e reactor coolsnt temperature anc oressure and system heatuo anc cooldo.n rates (with the orieption of the pressu*irer) shall be limited in accoroance with Figures 3.4 ~, and 3.4-3 for the service period specified thereon: .dlowable t.ombinations of pressure and temperature for specific a. temperature change rates are below And to the right of the limit line; $hown. Limit lines for cooldown rates between those presentta may be obtained by interpolation; and Figures 3.4 2 and 3.4 3 define limits to assure prevention of b. non ductile failure onl Characteristics, e.g., y. For normal operation, other inherent plant pump heat addition and pressurite" neater capacity, may limit the hestup and cooldown rates that can be achievec over certain pressure temperature ranges. 2. These 'imit l*$ shall te calculated periodica1E using methods orovided belo' 3, The secondary side of the steam generator must n:. te pressurized above 200 psig if the temoerature of the steam generate is below 70'F, 1. The prRssuriZer neatup and Cooldown rates shall net exceed 100*F/h and 200'Fih, respectively, and 5. System preservice hydrotests and inservice leak anc hydrotests sna11 be

erfnrmed at pressures in accordance with the requirements of ASME Boiler and Pressure vessel Code, Section x!.

The new 10 CFR 50, Appendix G rule addresses the retal temperature of the closure head flange and vessel flange regions. This rule states that the mini-mum retal temperature of the closure flange region should bJ at least 120'F higner than the limiting RT for these regions when the pressure exceeds 20% NDT of the preservice hydrostat,1c test pressure (621 psig for Westinghouse plam.s), f f4rde*6ac" %:w 'Ja't%/ne minimum temperature of the closure flancs and the vessel flange regions in 160'F since the linating RT is 40*F (see Taele B I NOT ) 3/4.4-1). The C: i m : S e u n ---I heatup and cooldown curves shown in Figures 3.4-2 a1d 3.4-3 are impacted by this new rule. COMANCHE PEAK - UNIT 1 B 3/4 4-6

3 ". 7 E ;: 7 r,.o 12-C)

i "J E?s

-4 I ABi l 11 3/4 4 - l a,. y uBT1 RiACIOR VISSTI I RA_CIllR_ _I10_118JiNI % PROP 3 RIIES = 50 f I-I 8 AVG Stil l a Alic. Sill !! I 35 Mit Ri INIRGY I NI RGY g, g Code Cu Ni P I[MP. WT){l.) NMwu(( ) _E COMPONINI t,wi& NO T. 1..

r I_ I - I.H I._I. I_a _ _
I

-4 C losur e 151. Dose A5338, Cl.I Ril10-I .les .61 .01/ fo 100 40 1/n.0 i Closure list. forus A5338, C1.I RIIIl-1 . 0 71 6I .sHIH 30 3:1 -30 1Ih 5 i Closure lid. I lange A508 Cl.2 R1102-1 .// .084 40 100 40 119 a Vessel Ilange A508 C1.2 Rilot-1 . /2 .011 10 10 10 91 o Inlet Nortle A508 Cl.2 RIIUS-1 .09 .H2 .010 -10 50 -10 14 / ti inlet Nortle A508 Cl.2 R1105-2 .11 .81 .011 -lil 40 -20 I th 5 Inlet Nutzte A508 Cl.2 R1105-3 .31 .81 012 -10 50 -10 114 0 Inlet Nozzle A508 Cl.2 RIIUS-4 .09 .87 .011 -10 50 -10 1% 5 R Outlet Hozzle A5C8 Cl.2 RIl06 1 .68 .004 -20 40 -20 115 0 Outlet Nortle A508 C1.2 RilD6-2 . te? .008 -10 50 -10 111 0 ? Outlet Nozzle A508 Cf.2 RI106-3 .64 .005 -20 50 -10 l15 5 Outlet Nozzle A508 C1.2 R1106-4 .65 .004 -20 40 -20 11/.5 Opper Shell A533B, Cl.1 kl104-1 .0/ .61 .012 -30 100 40 til u Upper $liell A5338, Cl.1 RilO4-2 .08 .6/ .011 -50 100 40 15 0 upper Shell A5338. C1.1 RIl04-3 .05 .60 .010 -20 10 10 101.5 Inter Shell A5338, Cl.1 R110/-l .06 .65 .010 -20 10 10 111.5 93.5 Inter SheII A5338 Cl.I Ril0/-2 .n6 s.1 .010 -23 50 -10 !?3 5 103 6 Inter $liell A5338, Cl.1 RIl0/-3 . t:5 .sa. .001 -20 10 10 131.0 BM o lower Stiell A5338 C1.1 RIIH8-1 .08 .64 006 -20 60 0 II9.o 85 0 tower Stiell A5338 C1.1 RIl08-2 05 .59 .006 -30 80 20 124.5 /H H tower-ShcIl A5338, C1.1 RIID8-3 .0/ .64 .008 -30 60 0 122.0 wi 0 Buttom Ikl. Iorus A533ff, CI.I R1I12-I I3 .62 .010 -50 50 -10 112.O Bottom lid Dome A5338 Cl.1 Rill 3-1 . 0!! .60 .010 -50 10 10 90.0 Int er. & t ower A5338, Cl.1 Gl.6/ .04 II .008 - 10 -10 -10 15n 0 shcIl (long. & Giv th WeId Seams)( s) a) B4 Weld Wire til 88112 & 1inde 00'11 Ilua lot No. 0115 te) Major Work i ng D i s ec t ion (l ong i t o.f i na l ) e) Norm.sl to Major Wese k ing Diret t oon (Iransverse)

W 12-C RC .gg -fA6LE 8 3H.4-Is qb5 f 'Z"Th' PEMUNIIM REACIOR VESSEL FRACTURE TOUGWlESS POOPERTIE { ~ 'XI ft-1b Avg. Wif Ene 35 Mil Cu Ni P NDI Temp. NOT MMWD NI ICI Component Code No. Grade

F
F
F ft-lb fL-MWG Closure Head Dome R3811-1 A53.B. C1. I 0.15

.65 0.014 -40 60 0 131 Closure Head Torus R3810-1 A5338 CI. 1 0.15 .69 0.011 -30 30 -30 143 Closure Head Flange R3802-1 A508, C1. 2 .71 0.013 40 ' <100 40 152 Ves el flange R3801-1 A508. C1. 2 .70 0.009 -10 <50 -10 121 Inlet Nozzle R3803-1 A508 C1. 2 .84 0.009 -10 <50 -10 138 Inlet Nozzle R3803-2 A508, C1. 2 0.10 .91 0.008 -20 <40 -20 13b Inlet Nozzle R3803-3 A508. C1. 2 .91 0.010 -10 <50 -10 146 Inlet Nozzle R3803-4 A508. C1. 2 .86 0.009 -20 <40 -20 136 Outlet Nozzle R3805-1 A508 C1. 2 .64 0.006 0 <60 0 132 Outlet Nozzle R3805-2 A508. C1. 2 .66 0.005 0 <60 0 119 i Dutlet Nozzle R3805-3 A508. C1. 2 .66 0.004 0 <60 0 117 N Outlet Nozzle R3805-4 A508. C1. 2 .67 0.005 0 <60 0 119 D Nozzle h il R3806-1 A5338, C1. 1 0.05 .61 0.010 -10 100 40 76 Nozzle Shell R3806-2 A5338, C1. 1 0.06 .62 0.009 -30 70 10 87 4 a, Nozzle Shell R3806-3 A5338, C1. 1 0.06 .70 0.007 -30 100 40 86 P Inter. h il R3807-1 A5338, C1. 1 0.06 .64 0.006 -20 <40 -20 108 133 i

nter. Shell R3807-2 A5338, Cl. 1 0.06

.64 0.007 -20 70 10 101 122 ' Inter. h il R3807-3 A5338, C1. 1 0.05 .60 0.007 -20 40 -20 105 120 tower Shell R3816-1 A5338, C1. 1 0.05 .59 0.001 -30 30 -30 107 136 Lower h l! R3816-2 A5338, C1. 1 0.03 .65 0.002 -30 60 0 106 131 Lower Shell R3816-3 A5338 C1. 1 0.04 .63 0.008 -40 20 -40 108 139 Bottom llead Torus R3813-1 A5338, C1. 1 0.12 .65 0.009 -60 0 -60 123 Bottom Head Dome R3814-1 A5338, Cl. 1 0.12 .56 0.009 -70 -10 -70 112 Weld Metal (a) 0.05 .03 0.004 -60 <0 -60 96 i (Inter. to Lower h il Girth Seae) Neld Metal (b) 0.07 .05 0.005 -50 <10 -50 772 (Inter. & Lower Shell Long Seams) (a) 84 Weld Wire Mt. 89833 & Linde 124 Flux Lot No. 1061 (b) B4 Weld Wire Mt. 89833 & Linde 0091 Flus Lot No. 1054 (c) -Normal to major working direction (d) Najor working direction - =.

Txx.9?001 ATTACHMENT 12 PAGE70F;{...;...;q. ! alii

!!*.:! *!ucrea*;:q.,;v; 3 (; -t<

,e:) -!a* D .e:;

e.se :' ; e ::me: site :seve 's ae:essary to set :enseevat*,e est ;

,initati:ns oe:ause it is 00ls' ole for ::noit'ons to exist such inat : e* t e

srse c' r e aest.o caec t"e c:atrolling c:rcition soit:nes ' rem t e

e t; t*e Outsi:e an: t*e Oressure 'imit t+st at all times ce caseo :n 3 s',s4 5

f tre most critical Oriterion.

The aew 10 CFA 50 A :encia G rule addresses the metal temceratsre Of tae D

losure read flange and vessel flange regions.

This rule states tnat the mint-tum tetal tem:eratsre of t*e closure flange region should be at least 120 :egrees-

nigree tNan t*e iimitd g RT for these *egions when the pressure excee:S yg7 23 per:ent of the Ortservice nydrostaji: test pressure (621 psig for 'aes',inga. esse 10 ants). he Ge*e***+ I4a t U^'t 17%ne minimum temperature of the citsste 31
1ange anc sessel 'lange regions is 160 cegrees-F since the limitiag R7qg7 is 40 degrees * (see Table B 3/4.4 1).

The h*sae4-he-eNIcelce.n

Lrves sne.n in rigsre 3.4-3 are itcactes :y tnis new rule, and tae ef:re the

'noten" in tne cecidc.n curves. Finat'y, t*e ::mcosite curses for tre heatvo rate cata and the ::::::.n rate cata are a justea for possitie errors in tre pres sre ana tetperat.re seasing 'estrutents by the values inoicatea on the ret:ective curves. Altnough t*e pressuri er coerates in temoerature 'snges above tacte f:r .n':n there is reas n for concern of nonductile f ailut e, operating limits are proviced to assure compatibility of operation with the fatigue analysis

erformee in at:oreance with the ASME Code requirements.

_CW *EMDE0ATURE OVERPRESSURF PROTECTION The CDERABILITY of two PORVs, two RHR suction relisf valves, or an RC5 vent ocening of at least 2.98 square inches ensures tnat the RCS will be protectee from prenture transients which could exceed the limits of 10 CFR 50 Appendix G .rer, one or more of the RCS cold legs are less than or eoval to 350"F. Eitner PORV or either RHR relief valve has adequate relieving ca:acility to protect tse 4C5 from overpressurization nen the transient is limited to either: (1) the start of an idle RCP with the secondary water temperature of the steam generator less tnan or equal to SC*F above the RCS cold leg. temperatures, or (2) tne start of two,crarging pumps and their injection into a water solid RCS. The maximum hominal Allowed PORV Setroint curve is derived from analyses .nich model the performance of the overpressure protection system for a range of mass input and heat input transients. Figure 3.4 4 is based upon nis analysis including consicerat', n of the maximum Dressure overshoot Devond the FORV set:oint anien can occur as a result of time delays in' signal

essing i

CCMANCHE :EAK - UNIT 1 B 3/4 4-13

TXX 'JE001 j- -ATTACHMENT 13 PAGE 1 0F 3 i i {_ SECTION: 3/4,f:.1 and BASES 3/4.6.1 c M ffAjLi JUSTirlCATIOff i i t 13 A 3/4 6 3 The valve identification numbers (tag numbers) contain j B 3/4 6 1 a unit designator as the first character, i.e.1HV 4776 would be a Unit i valve with 2HV 4776 being the i corresponding Unit 2 valve. The dual unit Technical 1 Specifications will utilize a convention of identifying i valves; without the unit designator if the remainder of i the tag number is applicable to both units, with the unit-designator if the tag is only applicable to one unit. 1 4 i (See revised BASES for specification 3.0.S). i' l f F b 'N 4

\\ TXX-92001 ATTACHMENT 13 : NTA!N9!N* !d*!us PAGE 2 0F 3 $URVEfLLANCE RE0V!8EutNil (Continued) If any periodic Iype A tatt f ails to meet either 0.75 L, or 0.75 L, b. g the test schacute for subsequent Type A tests shall be reviewee and accreveg tj the Commission. If two consecutive Type A tests 'til to meet ei eer !.*5 L, or 0.75 L, a Type A test shall be performee at t g ' east ever

! months until two consecutive Type A testl meet either 0.75., ;r,0.75 i at anich time the above test schecule may te resweeo; g

The a:curac, of each Type A test shall be verified by a supplemental c. test anich: 1) Confirms the accurat.y of the test by verifying that the succle-sental test result, L. is in accordance with the appropriate g fo11 ewing equation: IL - (L

'o) l 1 a or i L g

am c tm o) I 1 0.25 L g where L,, or L, is the measured Type A test leakage and L g is the superimposed leak; g 2) Has a duration sufficient to establish accurately the change in 'etrage este between the Type A test and the supplemental test; ano 3) Recui es that the rate at which gas is injected into the coatain-ment or eled from the containment during the supplemental test is :etaten 0.75 L, and 1.25 L,; or 0.75 L and 1.25 L. t t

ype B and C t O.

less than 3,. [3sts shall be conducted with :,a at a pressure act 48.]psig, at intervals no gre 4ter than 24 months except for tests involving: .) 41* 1:cks, 2) Coatainment ventilation isolation valves with resilient material set:1, 3) S a f e t;. injectien valves as specified in Specification 4.6.1.2g, anC 4)

ntainment spray valves as specifiec in Snecificatien 4.6.1.2n.

e. Air 'ccAs small be tested and demonstratec .a8LE by the recuire-ments of $cecification 4.6.1.3; Containmen* ventilation isolation valves with resilient materis' seals shall be testec and comonstratto OPERABLE ey the recuirements of Speci'ication 4.6.1.7.2 or 4.6.1,7.3, as applicable; g. Safety i de:ti:n valves @ 809A @ S09B. ano G8840 shall te 'ea testec t* s gas at a pressure net less than P,@8.3,]osio, or ith r, 13 A) ater at 3 a : essure not less than 1.1 P,, at intervals no ;-eater than 24 moatns;

entainmeat sceay valves @1V-4776, @V 4777, QCT-142, anc @T*'45 shall N

e les* teste:.ith wate* at a pressure not less than 1.1 P 5t ntersais

greater tnan 24 months; ana 3

- e :rasis :rs :f iceci'4:ation 4 0.2 are not applicaele.

Ma',;-E :E:..
M * '.

3/4 6-3 i

TXX 92001 ATTACHMENT 13 PAGE 3l0F "$ 3/4.6 00NTa!sWENT 5'5TEWS SA5E3 3/4 6 1 001Ma0y CONTA!NV[NT 3 /4. 6.1.1 CCNtaINutNT INTEcst'y Primary CONTAINMENT INTEGRITY ensures tnat the release of racioactive materials from the containment atmosphere will be restricted to those leakage paths ano associated leak rates assumed in the safety analyses. This restric-tion, in conjunction with the leanage rate limitation, will limit the ExCLUSICN AREA BOUNDARY raciation doses to within the cose guideline values of 10 CFR 100 during accident conditions, 3/4 6.1.2 CONTAINMENT LEAKAGE The limitations on containment leakage rates ensure that the total containment leakage volume will not exceed the value assumed in the safety analyses at the peak accident Dressure, P,. As an added conservatism, the ~ ~ measured overall integrated 1eakage rate is further limited to less than or equal to 0.75 L, or 0.75 L. as applicable, during performance of the periodic t test to account for possible degradation of the containment leakage barriers -tetween leakage tests. For specific system configurations, credit may be taken for a 30 day -ater-seal that_ *ill ce maintained to prevent containment a *csphere leakage through the penetraticns to the environment. The following iJ a list of the containment isolation _ valves that meet this system configuration 1 :: the Maximum Allowec Leakage Rate (MALR) required to maintain the water su' for 30 days. WALR Valve No. (cc/hr) SO9A "77' 809B 77 3 840 2577 13* A t T 142 4734 T-145 4734 V-4776 4734 V 4777 4734 The surveillance testing for measuring leakage rates is consistent with the requirements of-10 CFR 50 Appencia,1 3/4.6.1.3 CCNTA!NMENT AIR LOCKS The limitations on closure and leak rate for the containment air locks are-required to. meet the restrictions on CONTAINMENT =lNTEGRITY and. containment leak rate. Surveillance testing-of the air lock seals provides assurance that the overall air lock leakage *ill not become excessive due to seal damage curing tne intervals between air lock leanage tests. CCMANCHE ;!AK - UNIT 1 B 3/4 6-1 i

s TXX 92001 ATTACHMEtiT 14 PAGE 1 Or 2 i SECTION: 3/4.7.1 LTJB PAGE # JUSTl Flf611Rfl 14 A 3/4 7 2 The valve identification-numbers (tag numbers) contain a unit designator s the first character, i.e. 1HS 021 would be a Unit i valve with 2HS 021 being tne corrosponding Unit 2 valve. The dual unit Technical Specifications will utilire a convention of identifying valves; without the unit designator if the remainder of the tag number is applicable to both units, with the unit . designator if the tag is only applicable to one unit. (See revised BASES for specification 3.0.5). t F l

TXX-92001 i ATTACHMENT 14 PAGE 2 0F 2 i 3,, WAXlMUM A1.t.0WABLE POWER RANGE NEUTRON FLUX HIGH $ETPQ INOPERABLE STEAM LINE 5AFEh VALVES MAK! MUM NUMBER OF INOPERABLE max! MUM ALLOWABLE power RANGE $AFEtt VALVE 5 ON ANY OCERAT NG St!AM GENERATOR NEUTR(N FLVX HIGH $ETPOINT (DER 0ENT Or RATED TWERMAL COWED) 1 87 2 65 3 43 TABLE 3.7 2 STEAM LINE SAF!'Y VALVES PER LOOP VALVE NUMBER _LIrf SETTING (* 1%)* ORIFICE 5!!! LOOP 1 LOOP 2 LOOP 3 LOOP 4 GHS-021, 058.

093, 129 1185 psi; 16 in?

(#45-022, 059,

094, 130 1195 psig 16 in 2 H A! @45-023, s'

060, 095, 131 1205 psig 16 in 2 (D45 024, 061,

096, 132 1215 psig 16 in

@45-025, 062,

097, 133 1235 psig 16 in' S

'The lift setting pressure snail corresponc to ambient conditiens of tne valve at nominal operating temperat,are and pressure. COMANCHE ;EAK s UNIT 1 3/4 7-2 1

~.__ _ _ _ _. _ _ __. _ _ _ _.. ~.__.____- .-----._..____._m_.___, t I 1 TXX-92001-ATTACHMENT 15 PAGE I 0F 2 i SECTION: 3/4.7.5 4 Lild PAGE # JUST!FICATIOJJ f i 154A 3/4 7 15 The ultimate heat sink is common to both units et CPSES. The added notation i s an operational enhancement to identify that both units-shall comply with the required ACTION. (See proposed specification 3.0.Sa). i i i + I 5 t -r -.mv+ w-w vpe----"v M -42ywwT' '"'-YT'"*-*^*"**'*"-*-'***""~' T '*T'=v-- r-* v tm-'ee+-w'w+' "vw-r-- -$-'m ee*av N-m

+er ve=*-ar-+e*e-'

-=-*--==e--==-

TxX-92001 ATTACHMENT 15 PAGE 2 0F 2 DLANT $v$ fu$ 3/a 7.5 ULTIMATE HEAT $1NK LIMITING CCNDITICN FOR ODERATION

3. L 5 T~e M td tste etat link (VHS) small te C#ERABLE witm A minimum water lesel at or above elevation 770 feet Mean Sea Level.

4. V5GS datum, D. A ation service water intake temperature of less than or equal to [10 F, and A maximum average sediment depth of less than or equal to 1.5 feet c. in the service eater intake channel. ADDLICABILI?v: MODES 1, 2, 3, and A. ,,rrn - (iO; s:cN: :( a n m 1 -,t 2 ) j ~ ^ 'with the' ate've~feQu1riments for.ater level and intake temperature not a. satisfied, be in at least HOT STANDBY within 6 hours and in COLD SHUTOCWN .ithin the following 30 nours. D. With_the average sediment depth in the service w.* er intake channel greater tnan 1.5 feet, prepare and submit to the.;mmission within 30 cays, pursuant to Specification 6.9.2, a Special Repor*'. hat provices a recccc of all surveillances performed pursuant to Speci*' cation 4.7.5c anc specify *nat.teasures will be employed to remove sediment from tae service water intake channel. SURVE!LLANCE REOUIRfuENTS 4.7.5 The ultimate heat siak shall te cetermined OPERAELE: At least once per 24 hours by verifying the station service water a. intake temperature and UH5 water level to ce.ithin tneir limits, t. At least once per 12 months by visually inspecting the cam anc .erifying no abnormal cegracation or erosion, and At least once per 12 months by verifying that the average seciment c. cepth in the service water intake channel is less than or equal to 1.5 feet. COMANC-E ;EAN - UNIT 1 3/4 7-15

9 1xX 92001 ATTACHMENT 16 J. PAGE 1 0F 2 i l SECT 10th 3/4.7.6 j ~ 1]JJj PAGE # JU$7 ] f ] C AT ![Ly i 16 A 3/4 7 16 Flood Protection it, common to both units at CPSES. The added notation is an operational enhancement to identify that both units shall comply With the required ACTION. (See proposed specification 3.0.Sa). h ~ l-l l l

TAA-WtuVl ATTACHMENT 16 ptast !<5?EM5 .PAGE 2 0F 2 3/4 7,6 FLOOD 800TECT!0N t!w! TING CONDITION FOR OPERATICN i ),7,6 F1000 protection shall be Drovided for all safety *relateg systems, t ccmoonents, and strwetures unen tre water level of the $cuaw Creek Reservoir (SCR) exceecs 777.5 feet Mean Sea Level. USGS catum. AdPLICABIt!TY: at 31,1 time [1__, v -, ~ Ib'

E'IO";

Unifs ] aru) 2 With the waWr-1Wiht-SCR a6v)e elevation 777.5 fee datum, initiate and comDiete within 2 hours, the flood protection measures verifying that any eQuiDment which is to be opened or is opene0 for maintenance is isolated from the SCR by isolation valves, or stop gates, or is at an elevation above 790 feet. SURVE!LLaNCE REQUIREMENTS

7.6 7be water-level of $CR shall be cetermined to be within the limits by:

Measurement at least once per 24 hours when -*e water level is below 4. elevation 776 feet Mean Sea Level, USGS dat,>. Measurement at least once per 2 hours when *. e water level is eQV61 b. to or above elevation 776 feet Mean Sea Leve'. USGS datum, and With the water level of SCR above 777.0 feet Mean Sea Level, USGS c. catum, verify floco protection measures are in effect by verifying once per 12 nours that flow paths from the SCR which are open for maintenance are isolated from the SCR by isolation valves, or stop gates, or are at an elevation above 790 feet, 9 -COMANCHE PEAK - UNIT 1 3/4 7 16

i i i i 4 l l Txx 92001 ATTACHMENT 17 PAGE 1 0F 2 $ECTION: 3/4.7.7 JT[B P_A,$[jl JUSTIFICATION 17 A 3/4 7+17 The Control Room HVAC System is common to both units at CPSES. The added notation is an operational enhancement i to identify that both units shall comply with the i required ACTION. l I (See proposed specification '.0.6a), a I i l l~ i i = i l -t I-i ie j-l i 1 6 i i i l [ ? h In t i 4 i b t s b l-1-

I Txx-92001 i ATTACH!iENT 17 PAGE 2 0F 2 8'.ast 5f5'!MS 3/4 7.7 CONTROL ROOM HVAC SY$ TEM LWITING CONDITION CCR OPERATION i

3. 7. 7 Two indecencert control room HVAC trains sna11 be OPERABLE.

1 APDLICABILITY: All MOOK..., 'CUON: wnel 2 n a MOD E 5 1, 2, Ta'n~cf Ti"~ ~ With one Control room HVAC train inoper4 Die, restore the inoperaDie train to OPERABLE status witnin 7 cays or be in at least HOT STANOBY witnin the next 6 hours and in COLO SHUTDOWN within the following 30 hours. MODES $ and 6: With one control room HVAC train inoperable, restore the inocersele 3, train to CDERABLE status within 7 days or initiate and maintain operation of tne remaining OPERABLE control room HVAC train in the emergency recirculation mode. With botn control room HVAC trains inoperaD o, or with the OE!RABLE D. i control 'com HVAC trains required to be in

e emergency recircula-tion'moce Dy ACTION a... not capable of beinc oowered by an OPERABLE emergency power source, suspend all operati: s involving CORE ALTERATICNS or positive reactivity changes.

$URVE!LLANCE REQUlt!"ENTS 4.7,7 Eacn contrc2 room HVAC train shall be demonstrated OPERABLE: At least once per 31 days on a STAGGERED TEST BA515 by initiating, a. from the control room,-flow through the HEPA filters and charcoal adsorbers and verifying that the train operates for_at least 10 continuous hours with the emergency pressuritation unit neaters operating; 4 COMANCME DEAK - UNii l' 3/a 7-17

TXX-92001 ATTACHMENT 18 PAGE 1 0F 2 SECTION: 3/4.7.10 1 TEM fML!! 1 Elf 1Gl[nj 18 A 3/4 7-24 Building is being made plural to clarif y the distinction between the Units 1 and 2 Containment and Safeguards Buildings. I

TX X-92001 ATTACHMENT 18 PACL 2 0F 2

ais! ! '-3 l

EEa 'Euegoaty;g uCN!?0 RING AREA vt. a i,ut u !uDER A RE L:v:' (*r! Nor a1 rence ai ":ac't4:as

: t': s 1.

Electet:a1 l a :: ::atroi Sc lairg 13; q Normal areas 1:a Centrol Rocm Sain sevel (El 230' 0") 50 1C4 Control Room tecnnical Svocort Area (E1. 840' 4') 4 1:.4 104 UPS/Batteay Recms 104 u3 Chiller Eouipment Areas 122 131 2. Fuel Building Normal Areas 104 131 Spent Fuel Pool Cooling Pump Rooms 122 131 m ) 3. ~, SafeguardsBuiloings) t Normal Areas 104 131 AFW, RHR, 51. Contain,aent $ pray Puro Rooms 122 131 RHR Valve and Valve Isolation Tank Rooms 131 RHR/Ci Heat Excnanger Rooms 131 Diesel Generator Ares 131 Diesel Generator Ecuipment Rooms 131 Oay Tank Room 131 4. Auxiliary Buil:ing i Normal Areas I ICA 131 CCW. CCP Pumo Rooms 122 131 CCW Heat Exchanger Area U2 131 CVCS Valve and valve Operating Rooms 122 131 Auxiliary Steam Orain Tank Equipment Room 122 131 waste Gas Tank Valve Operating Roem 122 131 5. Service Water Intake Structure 127 131 (j 6. Containment Building [ .s General Areas 120 129 Reactor Cavity Exhaust 150 190 CRCM Shroud Exhaust -163 172 u l CCMANChE PEAK UNIT 1 3/4 7-24

l 1XX-92001 i ATTACHHEtJT 19 PAGE 1 0F 2 Sect!ON: 3/4.7.11 l M MGL,j JUSTIr1 CATION 19 A 3/4 7 25 The UPS HVAC System is common. to both units at CPSES. .The added notation is an operational enhancement to identify that both units shall comply with the required ACTION. The term ' ACTION:* was an inadvertent omission f rom the specification which should be included for completeness. (See proposed spectftcation 3.0.Sa). i 9 e s ._ _. _,,___.,~~,_---~__,_.-_ _-.. - -.-

TXX-92001 A1TACHMENT 19 PAGE 2 0F 2 a. TNT i q !.";. 3'a 7 11 V:5 ava; :'5'E'4 00!#ATlNG . !MI': *.0 !!'.C I T * :H 208 : ERAT!!N

b. 7.11 7 0 4"ce:eecent UPS WAC trains inall te OPERABLE, jhA a;#'.!! ABI L:? t:

"CCES 1. 2. 3 and 4 i 7.iithonlyoneLP5HVACtrainOPERABLE.restoretheinocerselesystemto CPERAlli status within 7 days or be in at least MOT STANOBY within the rest 6 hours anc in CC=LD SMuiOOWN within the following 30 hours. 50tvEILLaNCE RE;UIREMENTS 1.7.11.1 Eacn u;5 *VA: train shall :e cemonstratec OPERABLE at iesst ence per 1B mentns ey: d Ver fyir.g tMt each UPS HvaC train st* ts automatically on a safety a. Injecticn test signal. b. ~!eri'ying that ea;n UPS HVAC train starts a. :matically en a ilackcut telt signal, J.7.*1.2 Es:n.'D5 %AC train snall te cemenstrated 0::.;;BLE at least or:e ter 31 ca,s 0;. itarting the non operating UP3 HVAC tra'n anc verifying tant t*e train :serates for at least i nour, ACTION : ( Unih .1 ud 2 )__) L b g 1

_.u_ TXX 92001 ATTACHMENT 20 PAGE : 0F 3 SECTION: 3/4.8.2 111]j ffilf JUSTIr! CAT 104 20 A 3/4 8 11 (Byron precedence) The Unit 2 component identification numbers have been added in a format desired by CP$CS. 1 mm i u um ip

Txx 92001 ATTACHMENT 20 i 3,.... PAGE 2 0F 3 2-8,,, I

DE#A'!NG y: u!' NG COND!*:0N FOR OPE 8A?!*N 2C is)

L !. 2.1 As a minim m, the following 0.C. electritti sources shall te CCERA2'.!: u A d t O p 'stTb k at jteri 4-sna F 1eert oae f s11 uttity ceara s e tT7c1atec ach oat' ,, arc D N / q D. 3, / B 125 *' 5 tat y rtleries JBT' & d e Bil!** 1eas pt .c st c a p a c i tpsf g e r a s s o cj,ttetr i t h e a c3,*a ery f v p ADDLICA2!LITY: MODES 1, 2. 3, and 4. I i ACT0N: i i With ;ne of the reovired battery trains anc/or recuirec full capacity chargers incoeracle, restore the inoccrtDie battery train anc/or recuired full ca acity caarger to OPERABLE status within 2 hours or ce in at least HOT STAN B'.itsin tre neat 6 nours are in COLD SWUTOOWN.ithin the follo.ing 30 hours. 5JRVE!LLANCE RE00l#E9ENT5 A. 5. 2.1 Eacn 125 '.' O.:. station cattery and cnarger 5: 1 be cenorstrate: OPERABLE: At least :nce :e* 7 cajs cy serifjing that: a. 1)

ne parameters in TaDie A.5 2 meet tre Category A linits. an:

2) The total tattery terminal voltage is greater than cc e: vat 1: 10B volts on float unarge. COMANC-E DEAA ',;N!i 1 3 A i-11

TXX 92001 ATTACHMENT 20 PAGE 3 Of 3 .INSE RT C a. Train A 125 volt D.C. Station Batteries BT1ED1 and BT1ED3 for Unit 1 (BT2ED1 and BT2ED3 for Unit 2) and at least one full capacity charger associated with each battery, and b. Train B 125 volt D.C. Station Batterles BT1ED2 and BT1ED4 for Unit 1 (BT2ED2 and BT2ED4 for Unit 2) and at least one full capactly charger associated with each battery. i --,,,--a,-g n e e i oy

TXX 92001 ATTACHMEtiT 21 PAGE 1 Of 4-SECT 10ft: 3/4.8.3 J.TJJj PAGE # 1USTIFICAT10ff 21 A 3/4 8 15 (Byron precedence) The Unit 2 cornperient idetitificatiori numbers have been added in a f ortnat desired by CI'5ES. = "we'-'PP-'WNw*7-MM TP -""'W9T-e F WNMMW4 P r=gg $ > W 'Tvrg7r-M-Tve*--gy'T33Fe& qMey

"'WW'T

'"V' Y"' G wet-b-bad w+4 y- +um--+eeMvs h(Y'maMe-- 3rg gM,,m.gwp-+y9 qw yg12 -prv*p-r"'tr'9*' -Y e#W'W r

.Tn-92001 ATTACHMENT 21 PAGE 2 0F 4 3 3 .*451*E 8%Et M SM!BLm 08EeAT!NG j .!w!?!NG 00N0!T!04 FOR CPERATION

3. 5. 3.1

- e folio.ing electrical busses sneli be energizac in tne scec m d e

anner: j AA.C.fergencyBusses :nsisti g of: a a. Trai [2q 1) 6900 vo t Emerg cy But EA1, L 2) 480 vo t Emerg cy Bus EB1 fr transf 'emer 11E 1, anc 3 480 v t Emer ncy Bu 1EB3 fr m trans ormer 71 83. b. Train B .C. Emer ency B ses con is f: Bus 1Eh, ting 1) 69 0 volt .ergene 2) Af0voltF ergenc Bus 1EM from t ansforme T1EC2, nc 3)

80* Volt Emergen Buslif4from raris form 71EB4 c.

11B ott A.. Instr entBus/IPC1an 1EC1 ene gized f om its as ociated nyerter connected to 0.. Bus 1ED *; r o c. B volt .C. Ins rument us 1PC2 nd 1EC2 nergite from 4ts associa o inver er conn tea to .C. Bus 02"; t W 118 Vo t A.C. ' strumerA Bus IP and 1EC energi ea from 4,6 assoc ated e. I inve er conn.ted to '.C. Bus E03*; h' 'le voit A,* Instru ent Bus PC4 and C6 4 er ited fee its as cistea in erter co.nected o 0.C. B s 1E04*: g. rain A 1 5-/olt .C. Buss s IED1 ar 1E03 e"ergized f em Sta on .atterie BT1EDI no Bi1E 3, respec ively; c n. Train 125 vol D.C. B sses 1E02 and 1E04 energize frem ation Batte ses BTIE 2 and B* ED4, res ectively a 8 8 '. :C AB !'. !'v : POCE3_1. 2, 3, and 4 .A_ ! 0 N : With one of the recuired trains of A.C. emergency ogsses cet fully 4. energized, reenergize the trains within 8 nours or be in at least HOT STAN0BY within the next 6 hours and in COLD SHUTOOWN within tee following 30 hours, i

ine inverters.may oe cisconnected from one O.C. bus f:e we to 24 neurs as necessary,. for the pur.ose of performing an ecualizing charge on thei; asso-ciated cattery train Leovideo:

(1) their instrument ousses are energf:ec. E [ and (2) tre-instrumeat :.sses associated witn the other cattery train are i energizec from tneir associatec inverters and connectes to their asso:iatec L D.C. bus. 0"MANCHE OEAK UNIT 1 1/4 2 15

TXX-92001 ATTACHMENT 21 PAGE 3 0F 4 USN g 6 t Train A A.C. Emergency Busses consisting of: a. 1) 6900 Volt Emergency Bus 1EA1 for Unit 1 (2EA1 for Unit 2), 2) 480 Volt Emergency Bus 1EB1 from f transformer TIEB1 for Unit 1 (2EB1 from transformer T2EB1 for Unit 2), and 3) 480 Volt Emergency Bus 1EB3 from transformer T1EB3 for Unit 1 (2EB3 from transformer T2EB3 for Unit 2), b. Train B A.C. Emergency Busses consisting of: 1) 6900 Volt Emergency Bus 1EA2 for Unit 1 (2EA2 for Unit 2), 2) 480 Volt Emergency Bus 1EB2 from transformer T1EB2 for Unit 1 (2EB2 from transformer-T2EB2 for Unit 2), and 3) 480 Volt Emergency-Bus 1EB4 from transformer T1EB4 for Unit 1 (2EB4 from transformer T2EB4 for Unit 2), c. 118 Volt A.C. atrument Bus 1PC1 and 1EC1 for Un:t 1 (2PC1 and 2EC1 for Unit 2) energized from its associated inverter connected to D.C. Bus 1EDi* for Unit 1 (2EDi* for Unit 2); d. 118 Volt A.C. Instrument Bus 1PC2 and 1EC2 for - Unit 1 (2PC2 and 2EC2 for-Unit 2) energized from its associated inverter connected to D.C. Bus 1ED2' for Unit 1 (2ED2*'for Unit 2); e. 118 Volt A.C. Instrument Bus 1PC3 and 1ECS for L Unit 1 (2PC3 and 2ECS for Unit 2) energized fron. its associated inverter connected to D.C. Bus 1ED3'. for Unit 1 (2ED3' for Unit 2); f. 118 Volt A.C. Instrument Bus 1PC4 and 1EC6 for Unit 1 (2PC4 and 2EC6 for Unit 2) energized from '6m'$ g'. g g-,..yw-wdmq v-w 'p y-a.g-y p pye*-- irme ywn.,, +- rgg-eg.y as se4-w+ veem,eewes eL----e-m-w -a --r -+smu-em e*ri y

Ihli$!kT21 TN5*RT O lC*d) .PAGE 4 Of 4 its associated inverter connected to D.C. Bus 1ED4' for Unit 1 (2ED4' for Unit 2); { g._ Train A 125 Volt D.C. Busses 1ED1 and 1ED3 for i Unit 1 (2ED1 and 2ED3 for Unit 2) energized from l Station Batteries BT1ED1 and BT1ED3 for Unit 1 l (BT2ED1 and BT2ED3 for Unit 2), respectively; and h. Train B 125 Volt D.C. Busses 1ED2 and 1ED4 for Unit 1 (2ED2 and PED 4 for Unit 2) energized from Station Batteries BT1ED2 and BTIED4 for Unit 1 (BT2ED2 and BT2ED4 for Unit 2), respectively. i h i L.-

i TXX 92001 i 1 ATTACHHf.NI 22 PAGE 1 0F 2 i P $ECTION: 3/4.9.1 IJJB ffaL.i JUSTiflCATION 22 A 3/4 9 1 The valve identification numbers (tag numbers) contain a unit designator as the first character, i.e,105 84% would be a Unit 1 valve with 2CS 84% being the corresponding Unit 2 valve. Ths dual unit Technical Specifications will utilize a convention of identifying valves; without the unit designator if the remainder of the tag number is applicable to both 9 nits, with the unit t designator if the tag is only applicable to one unit. (See rtvised BASE $ for specification 3.0.5). k I t r I I t y9--wwce-e-ag a % + r y W+%ew av.-wC---+++g-vegwee-s-m-eww-+ws+Wy.-- wpvvp-Q p m,9 ww --pnepewy.,n em.- v v pq_eg.%gq-em-w 9m-pe y.q.mg. g gy y p y 9 ww+ 9.wy+ ve -M q -ry-pm+w.n-Newrwdy-r--ppysy 32'g h w wg

TXX-92001 ATTACHMENT 22 PAGE 2 0F ?. 3,4 9 REFUELING CPf4at'CN$ 3/4.9.1 BORON CONCENTRATION L:MITING CONDIT1CN FOR OPERATION 3.9.1 The boron cencentration of all filled portions of the Reactor Coolant Syst;m and the refueling canal shall be maintained uniform and sufficient to -ensure that the more restrictive of the following reactivity concitions is met; e i,t he r: A K,ff of 0.95 or less, or a. A boren concentration of greater than or equal to 2000 ppm.* di additionally, either valve @5 8455 er valves @$ 8560. FCV 1118,(ftS 8439, @$ 8441 and @.5 8453 snail be closed and secured in position. d t APPLICABILITY: MODE 6. ACTION: i With the reovirements a. or b. of the above not satisfied, immediately a. suspend all operations involving CORE ALTERATIONS or positive reacti-i vity changes and initiate and continue boration at greater tnan or [ ecual to 30 gem of a solution containing greater than or toual to (22k 7000 ppm boren or its equivalent until K,gf i s reduced to less than \\ h or equal to 0.95 or the boron concentration is rettored to greater than or soual to 2000 ppm, whichever is the more restrictive. If either valve @S-8455 or valves $$ 8560. "CV 1118, @S-8439, 'k b. @C5-8441 and @$-8453 are not closed and sec. red in position, immediately suspend all operations involvinc CORE ALTERATIONS or positive reactivity changes and take action *o isolate the dilution paths. Within 1 hour, verify the more rest.tetive of 3.9.1.a or 3.9.1.0 or carry out Action a. above. 3 SURVEILLANCE RE0018EMENTS \\ g 4.i.1.1 The more restrictive of the above two reactivity conditions shall te l cetermined prior to Reme;ing er unbolting the reactor vessel head, and ~ 4. i b. Withdrawal of any control rod in excess of 3 feet from its fully inserted position within the reactor vessel. 4.9.1.2 The boron concentration of the Reactor Coolant System and the refueling canal shall be determined by chemical analysis at least once per 72 hours. 4.9.1.3 Eithervalve(85-8455 or valves @S 8560, FCV 111B,(JtS 8439,($tS 8441 and @ S 8453 shall be verified closed and secured in position by meenanical stops or by remove; of air or electrical power at least once per 31 days to verify that dilution paths are isolated. "During initial fuel i csc, tre boron concentration iimitation for the refueling canal is not acclica 1e orovioed the refueling canal level is verified to ce below the reacto m iel flange elevation at least once per 12 hours. COMANCHE PEAK - UNIT 1 3/4 9 1 W- -py-y 9-w-g--se w--yy + vvym e -evww my gy-v 9 pry -w g y ye e y y- -p.,wy r y, .-,9-w-,--g-+ ,-,,,m,,,.mmg,w-w -ewr -r-e-me---- m-i--eW--- 4'r -E-- PT-"'TN-"*"

TXX-92001 ATTACHMENT 23 FAGE 1 0F 3 SEC T 10ti: 3/4.11,2 .LTE1 P A G f._, l!,!ST li l C ATION 1 23-A 3/4 11 2 The WASTE GAS HOLDUP SYSTE't is common to both units at CPSES The added notation is an operationci enhancement to identify that both units shall cc.nply with the required ACTION. (See proposed specification 3.0.ta). 23 B 3/4 11-3 The gas storage tank s are common to both units at CPSES. The added notation is an operational enh m oment to identify that both units shall comply with th ' equired ACTION. tSee proposed specification 3 0.5a).

TXX-92001 ATTACHMENT 23 :::::AC :/E E88LLEN'S PAGE 2 0F 3 3c4.11.2 CA5E005'ErFLUINTS E*Du;51VE GAS MIXTURE

v!T

N3 ::NCI' 0N rCR CPERATION 3.11.2.;

Tne concentration of oxygen in the WASTE GAS HOLCUP 5< STEM saail ce limsteo to less tnan or coual to 3T, ey volume whenever the hydrogen concentrati:n. ex'ceers 4% by volume. APDLICABilllY: At all ti ts f ,,, y (23-/C 'C" UN: biYs 1W2 Tittet^n~e'co'ncentYaN5nofoxygenintheWASTEGASHOLOUPSiSTEM a. greater than 3% by volume but less tnan or equal to 4% by velute, reduce the oxygen concentration to the above limits within 48 nears. With the concentration of oxygen in the WASTE GAS HOLDUP SYSTEM D. greater than 4% by volute and the hydrogen concentration greater than A% by volume, immediately suspend all additions of waste gases to the system and reduce the concentration cf oxygen to less than or equal to 4% by volume, then take ACTION a,,

aDove, Tne provisions of Specifications 3.0.3 are not applicable.

c. SURVE LLANCE REQUIREMENT 5 4.11.2.1 The concentrations of nydrogen and oxygen i the WASTE GAS wCL')UF SYSTEM sr.all be determined to be within the above limits by continuously monitoring the waste gases in the WASTE GAS H3 LOUP SYSTEM with tne nyaregen and oxygen monitors required OPERABLE by Table 3,3-7 of Specificatien 3.3.3.4, or oy the associated ACTION statements. COMANCHE 3Eh - UNIT 1 3/4 11-2 l _ _, _ _ _ _, _ _ _ _ _ _ _ _ _ _ - - - - - - - - - - - - - - - ~ ' - - - - - - - - - - - -

TXXk92001-ATTACHMENT 23 PAGE-3 0F 3

A010 ACTIVE EnLUENTS GAS STORAGE-TANKS' LIMITING CONDITION FOR OPERATION 3.11.2.2 The asantity of radioactivity contained in each gas storage-tank snal) be limited to less-than or equal to 200,000 Curies of noble gases (gonsidered as <c-133 equivalent).

APDLICABILITY( At_,all-t4ges , ~. 4 B' ACTION: / ( W ; fy } p d 2 f' $1th't,h[ Ia b y U dioactive material in any gas storage tank a. exceeding the above limit, immediately suspend all additions of racioactive material to the tank, within 48 hours reduce the tank contents-to within the limit, and describe-the events leading to this condition in the next Semiannual Racioactive Efflu nt Release Report, . pursuant to Specification 6.9.1.4. b. The_ provisions of Specifiestions 3.0.3 are not apolicable. SURVFILLANCE REOUT.REMENTS 4.11.2.2 The quantity of radioactive material contained in each gas. storage tank shall be determined to be within the above limit at least once per 92 cays when radioactive materials are being added to the tank. 'l COMANCHE PEAK +- UNIT 1 3/4 11-3

t TXX 92001 ' ATTACHMENT 24-t 1 AGE 1 0F 2 SECTION: 6.1. JTiti PAGE # JUSTIFICATION 24 A 6-1 The term unit refers to only one of the nuclear I generation facilities. The Plant Manager is responsible for the operation of both units, 9 ,,. - ~ ~ ,a-+. + .n.. .---~,..,, -.

TXX-92001 ATTACHMENT 24- -PAGE 2 OF 2 p.,y33.;...,.g - y.: g 6.' 1 4ESPONSIBILITY

6. L 1 The Vice President, Nuclear C;erations shall te responsible <:r cweral!

oceration of the site, while the Plant Manager sn411 be responsi:le f:r oce s-(2 N ' sma:n of'the uni.. The vice President, Nuclear Ocerations anc tne Plant '1aaager it ll eacn cele j in writing the succession to tnis responsibilit/ owring t b-15eir aosence, j F. I. 2 The Snif tYuoervisor (or curing his absence f em tne control room, a

esignatec indivicual, see Table 6.2-1) shall te resoonsiele for tre c:ntrei

cm c mmenc function.

~ A management cirecthe to this effect, signeo y tae vice President, Nuclear Operations snall te reissued to all station :ersonnel on an annual basis. 6.2 ORGANIZATION

6. 2.1 ONSITE AND OFESITE ORGANIZATION A: onsite and an offsite organization snall be established for unit

. operation and corporate management, respectively. The onsite and offsite organization shall incluce the positions for activities affect g tne safety of the nuclear power p' ant, Lines of authority, responsibility and communicati_on snail ce n. estaelished and defined from the nignest management levels tnrougn intermeciate levels to and including all operating organization -positions. Those relationships snail be do: 9 entec anc uccatec, as approcriate, in the form of organizational arts, functional cescriptions of decartmental responsib.!liti-, and relationsnics, aac job cescriptions for key personnel ec.sitiers. or in the ecuivalent forms'of cocumentation. These recairements : 9411 De documentec in tne F5AR. The Vice Dresicent. Nucler.- Ooerations snall :e resocnsib!e 'or overall site safe coa etion and shall have control over tnese nsite activities necessery for safe coeration and maintenance of tne clant. Tne Executive Vice President Nuclear Engineecing and Operatiers c. snali have corporate responsibility for_overail plant nuclear safety and snall take any measures neeced to ensure acceptacle cerformance of the staf f in operating, maintaining, and revicing tecnndcal support to the plant to ensure nuclear safety. The-individuals who train the operating staff anc those wne car y c. out the radiation protection and quality assurance functions may report to the appropriate _ manager onsitei however, tney snail nave sufficient organizational freedom to ensure their inoeoencence 'aom c:erating pressures.

i. 2. 2 UNIT STAFF ine unit _organi:ation snall te subject to the following:

4 a. Eacn on-cuty snift snall ce comoosed of at least the minimum s-S crew composition snown in Tacle 6.2-1;

"ANCHE PEAK - UNIT 1 6-1

4 TXX 92001 ATTACHMENT 25- ' PAGE 1 0F 5' SECTION: 6.2 e JTW PAGE # JUSTIFICATION

t 25 A' 62 (McGuire precedence) The changes are consistent with the required Unit Staff for two units with a common control 4,

room, 1

25 8 6-3 The changes are consistent with the required Unit Staff for two units with a common-control room i n a manner + consistent with the Standard Technical Specifications. 4 i k i -- r-h: e 5 i A-1 i ,e n e- ..-e w .--w a------4y-

. -. - - - ~. ~ -. l ATTACHMENT'25-- 'PAGE 2 0F 5'- .;, g ;.u-*d STRMS ' STMF (Continued) ^ bC] he Noe ed uM C' At least one ITMinsed OperathrgrfstTt in_the ontrol room wren 5. %el is-in @ reactor. In accition, while unit is im MODE 1, g5-A l. _y, 3,,, 4, at least one licensec Senior Operator sna11 ce in the control room; i daciation Protection Tecnnician" and a Chemistry Technician

sh c,

te on site _when fuel is in the retetor; All CORE ALTERATIONS snail te coserved and directly supervisec c, either a licensec Senior Ocerator or licensed Senior Operator Limited to Fuel Hancling who nas no other Concurrent responsibilities curing this operation; A. site Fire-_ Brigade of at least five members" shall be maintainea on e, site at all times. .The Fire Brigace shall not incluce_the Shift Supervisor anc tne two<ctner memcers of the minimum shift crew necessary;for_ safe shutcown of the unit and any personnel requirec for other essential functions curing a fire emergency; f. Administrative crocedures snail be ceveloped and implemented to limit tne.ornin functions (e.g.,g nours of unit staff'who perform safety-elatec licensec 3enior Operators, licensed Operators, Raciation Protection Tecnnicians, auxiliary operators anc ney maintenance personnel). The amount of overtime workea ey unit-staff embers performing safety-related functions snail De limited i* sccorcance witn tne NRC Solicy Statement:en working nours (Generic.nter No . 82-12); anc Tre Shift Operations Manager-snail mold a Senior Reactor Coerator . g, license. "The Raciation Protection and the Chemistry __Tecnnicians and Fire Brigace composition may be less than the minimum requirements for a ceriod of time not to exceec 2 nours, in orcer to accommooate unexpected absence, provicec immeciate action is taken to fill the receired positions. COMANCHE 154 - UNIT 1 6-2 i {

L TXX-92001 _. ' ATTACHMENT 25 -PAGE 3 0F 5'. 78LE 6.2-1 MU!5H T EW MPO, TIk [ N 'ING(E MIT 9dCIL}iY / / / / / / L/ ' / :%m/' / #"S# of##5 VN#o ty*t'Vaos@L r.___.[' M0pd1/2,[or/4 [ [ f0E for j/ / SS ~ I 570 ~1 1 No 0 2 1 A0 2 57 1 .f 1 A* / / / / / / / / i i i i ~ .s A 55 i --i i Shif t 'Suoervisor with a Senior Operator licenselof>ArAt/9 s SRO --- Incividual witn a -Senior Operator license M t/nM 11 RO Incisicual witn an Operater license fih AM t(2 AO Auxiliary coerator STA Shift Tecnnical Advisor The snift :rew ccmocsition may be one less than the minimum requirements o Tacle 6.2-1 for a pedoa of time not to exceed 2 hours in order to ac:ommc unexcected absence of on-cuty shift crew members provided immeciate acti:n is taken to: restore tne snift crew composition to within the minimum requirements l-of Table 5.2-1. This :coeision coes not permit any shift crew position :: unmannec uoon shif t :nange due to an oncoming shif t c - man being late or ce absent. During ary aosence of tne Shift Supervisor from the c. trol room wnile tne un is in MO:E 1, 2, 3. : J. an incividual with a valic snail te cesignatec to assume tne control room commanc.,nior Operator license function. C E-5 or-6, ar 4 :itne ind't Sucervisor from the control room anile tne unitDuring an acserte Of is in li:erse snali ce ces';ratec to assume the control room-commano fu l n YNSERT"3 L S The STA position saal'_ce mannec in MODES 1, 2, 3, ano a unless tne ini't Suce-visor or tne 'nci.-4:ual.itn a Senior Ooerator license meets the cuali:ations ces:- :ec :n Oct on 1 of the Commission Policy Statemeat :n i Erg "ee-ng Excert se (50 F4 4362'. Cctener 28, 1985) i 20MANC'-E 2EA( uN:-.- 5-3

Txx-92001 ATTACHMENT 25 PAGE 4 0F 5 _TNSERT I MINIMUM SHIFT CREW COMPOSmON TWO UNITS WITH A COMMON CONTROL ROOM NUMBER OF INDMDUALS Rh0UIRED TO FILL POSmON POSITION BOTH UNITS IN BOTH UNITS IN ONE UNIT IN MODE 1,2,3, or 4 MODE 1,2,3, MODE 5 or 6 AND or 4 or DEFUELED ONE UNITIN MODE 5 or 6or DEFUELED SS 1 1 1 SRO 1 none ** 1 RO 3" 2* 3" AO 3" 3* 3" STA 1*** none 1***

lhh5!h!NT 25': .IN sER T I

PAGE 5 0F 5

At least one of the required individuals must be assigned to the designated

position for each unit.

' **At least one licensed Senior Operator or licensed Senior Operator Umited to - Fuel Handling must be present during CORE ALTERATIONS on either unit, who has no other concurrent responsibilities, i

.- -...... - - - - -... ~.. ~. . -. ~. -. ~.. w TXXs92001 1 ATTACHMENT 26-PAGE 1 0F 2 SECTION: 6.9 Illt! PAGE #_ JUSTIFICATION 26 A 6 17 The Startup Report is a unit specific document. The term plant is used in reference to Loth units at the same. time, thus necessitating the change to the term unit. P P K ( r + .,. ~.. - - -n. ,,n 4. .r,- ..,,m,v e

TXX-92001 ATTACHMENT 26- 'PAGE 2 0F.2 ADMIN!$TRATfvE CONTROLS PROCEDURES AND PROGRAMS (Continued) Radioactive Effluent controls Procram (Continued) e.

10) Limitations on the annual dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of adioactivity and to radiation from uranium fuel cycle sources conforming to 40 CFR 190.

f. Radiological Environmental Monitorino Program A program shall be provided to monitor the radiation and radionuclides in the environs of. the plant. The program shall provide (1) repre-sentative measurements of radioactivity in the highest potential exposure pathways, and (2) verification of the accuracy of the effluent monitoring program and modeling of environmental exposure pathways. The program shall (1) be contained in the 00CM, (2) conform to the guidance of Appendix I to 10 CFR 50, and (3) include the following: 1) Monitoring sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance witn tne methodology and parameters in the 00CM, 2) A Land Use Census to ensure that changes in the use of areas at and beyond the SITE BOUNCARY are identified and that modifications to the monitoring program are made if qquired by the results of this census, and 3) Participation in a Interlaooratory Com.trison Program to ensure that incependent checks in the precisi: and accuracy of tne measurements of radioactive materials ', environmental sample matrices are performed as part of the quality assurance program for environmental monitoring. 6.9 REo0RTING REQUIREMENTS ROUTINE REPORTS-6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be suomitted to the Regional Administrator of the Regional Office of the NRC unless otnerwise noted. STARTUP REPORT ]/~[' 6. 9.1.1 unit A summary report of EdB startup and power escalatw 'esting shall-ce saemitted following: (1) receipt of an Operating License, is ' amencment to the license involving a planned increase in power level (3) in:tailation of fuel that has a different design or has oeen manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the unit. COMANCHE PEAK' UNIT 1 6-17

. - _ _ ~ -...,.. -.. -.. i i i TXX-92001' ATTACHMENT 27. PAGE-4'of 41 ? t. Justification The specifications in this attachment contain plant specific numbers which are - yet to be confirmed. Those numbers still requiring design confirmation, of applicability for Unit 2, have been enclosed within brackets for identification

purposes, t

E. 4 4 l 4 6 1 d. 2 9. .x N + 1 f ,7y.- m-5g-.*y-

myyy, y

C w T7- * ' '""

3MM. ' TABLE'2.2-1 (Continued) ' %YN + ~ 2 2L.' i

I O REACIOR TRIP SYSTEN'INSTRUNENTATION TRIP SETPOINTS ls E!3T jg TOTAL-SENSOR

" E$ - r ALLOWANCE ERROR l ,gl ' FUNCTIONAL UNIT (TA) Z ~(S) TRIP SETPOINT ALLOWABLE VALUE s 18. Reactor Trip System Interlocks c .z II, a. Intermediate Range

N.A.

N.A. N.A. I x-10 88 mg>s 16 x 10 88 amps H Neutron' Flux, P-6 b. Low Power Reactor Trips 'I Block, P-7

1) P-10 input N.A.

N.A. N.A. 10% of RIP * .512.7% of RIP *

2) P-13 input N.A.

N.A. N.A. [15h RTP* Turbine $23%RTP*' Turbine 'F(irstStagePressure I q) First Stage p s-c' sure Equivalent Equivalent c. Power Range Neutron N.A. N.A. N'. A. 48% of RTP* 150.7% of RIP * { Flux, P-8 d. Power. Range Neutron N.A. N.A. -N.A. 1 2.7% of RTP* 1 0% of RTP* 5 5 Flux, P-9 e. Power Range Neutron M.A. N.A. N.A. 10% of RTP* 17.3% of RTP* I Flux, P-10 + i 19. Reactor Trip Breakers M.A. N.A. N.A N.A. N.A. 20. Automatic Trip and Interlock N.A. N.A. N.A. N.A. N.A. -{ Logic i

RIP = RAIED.IHERMAL POWER

.. ~,

TXX-92001 ATTACHMENT 27 lPAGE 3 0F 41 3 1 EA:*: :*. C:N';0L 5 5'EuS 3'4.1.1 BORATION' CONTROL SwuTOOWN MARGIN T, g GREATED 'uAN 200'c

M!?:N3 :0N0!T::N OR OPERATION 3.1.1.1 Tre SHUTOOWN MARGIN snali ce greater than or equal to 1.6% ik/5.

APPLICABILITY: M00E5 1, 2". 3, ana 4 ACTION: witn the SHUTDOWN MARGIN less than[1.6h Ak/k, immediately initiate and con-tinue coration at greater than or equal to 30 gem of a solution containing greater than or equal to 7,000 ppm co on or equivalent until tne reovired SMUT 00WN MARGIN is restored. SURVEILLANCE REOUIREuENTS 4.1.1.1.1 The SHUT 00WN MARGIN shall De cetermined to be greater than or equal to [1.hi ak/k: a. Witnin 1 hour after detection of an inopera, e control roc (s) 1nd at least once per 12 hours thereafter anile e roc (s) is inoperacle. If tne inoceracle control roc is immovaole untrippaDie, the acove recuirea SHUTDOWN MARGIN shall De verified + :eptacle with an ircrease allowarce 'cr the withcra.n worth of tne in-.ac'e or untrippaole -control roc (s); ahen in "0CE 1 or M0 E 2 witn h,ff great.cc taan or equal to ' at b. least once per 12 nours oy verifying tnat cc-t-:1 cank.itt.cra-al is -itnin the limits of Scecification 3.1.3.5; c. 'anen in MODE 2 with K less than 1, witnin A nours prior.to eff achieving reactor criticality by verifying : at the predicted. critical contro'

d ocsition is witnin t%e i'mits of $cecification 3,1.3.6; o ior to initial coerati:n acove 54 RATED THEiyAL POWER af ttc each d.

r f uel loacirg, Of consideration of tne f actors f Speci'ic? - tion 4.1.1. i. le. e t o*, ith tre centrol-cants at tne max mur inst '- tion limit of spe:ification 1.1.3 6; and "See Soecial Test Exceotiens Soecsfication 3.10.1.

MAN;-E EAs - '.' N : ' 1 3*a 1-1

TXX-92001 ATTACHMENT 27 PAGE 4 0F 41 REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN - T,y LESS M M Em TO 20W LIMITING CONDITION FOR OPEkATION 3.1.1.2 The SHUTDOWN MARGIN shall be greeter than or equal to[1.2 ok/k. APPLICABILITY: MODE 5. ACTION: With the SHUTDOWN MARGIN less than[1.51f, ak/k, immediately initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7,000 ppm boron or equivalent until the required SHUTOOWN MARGIN is restored. SURVEILLANCE REQUIREMENTS = 4.1.1.2 to[1.3% Ak/k:The SHUTDOWN MARGIN shall be determined to be greater than o Within 1 hour after detection of an inoperable control rod (s) and a. at least once per 12 hours thereafter while the rod (s) is inoperable. If the inoperable control rod is immovable or untrippable, the SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod (s); and b. At least once per 24 hours by consideration of the following factors: 1) Reactor Coolant System boron concentration, 2) Control rod position, 3) Reactor Coolant System average temperature, d) Fuel burnup based on gross thermal energy generation, 5) Xenon concentration, and 6) Samarium concentration. l COMANCHE PEAK - UNIT 1 3/4 1-3 Amendment No. 5

TXX-92001 ATTACHf1ENT 27 PAGE 5 0F 41 REACTIVITY CONTROL SYSTEMS FLOW PATHS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.2 At least two of the following three boron injection flow paths shall be OPERABLE: The flow path from the boric acid storage tanks via either a boric a. acid transfer pump or a gravity feed connection and a charging pump to the Reactor Coolant System (RCS), and b. Two flow paths from the refueling water storage tank via centrifugal charging pumps to the RCS. . APPLICATION: H0 DES 1, 2, 3, and 4.* ACTION: With only one of the above required boron injection flow paths to the RCS OPERABLE, restore at least two boron injection flow paths to the RCS to OPERABLE status within 72 hours or be in at east HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least .jkak/kat200*Fwithinthe di next 6 hours; restore at least two flow pat s to OPERABLE status within the next 7 days or be in COLD SHUTDOWN sithin the next 30 hours. SURVEILLANCE REQUIREMENTS 4.1.2.2 At least two of the above required flow paths shall be demonstrated CPERABLE: At least once per 7 days by verifying that the temperature of the a. flow path from the boric acid storage tanks is greater than or equal to 65*F when it is a required water source; b. At least once per 31 days by verifying that oach valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position; and At 16ast once per 18 months by verifying that the flow path required c. by Specification 3.1.2.2a. delivers at least 30 gpm to the RCS. "A maximum of two charging pumps shall be OPERABLE whenever the temperature of one or more of the RCS cold legs is less than or equal to 350*F except when Specification 3.4.8.3 is not applicable. An inoperable pump may be e ergized for testing provided the discharge of the pump has been isolated from the RCS by a closed isolation valve (s) with power removed from the valve operator (s) or by a manual isolation valve (s) secured in the closed position. COMANCHE PEAK - UNIT 1 3/4 1-8 Amendment No. 5

TX'X-92001 ATTACHMENT 27 -PAGE 6 0F 41 REACTIVITY CONTROL SYSTEMS CHARGING PUMPS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.4 At least two centrifugal charging pumps shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3*, and 4** *. ACTION: With only one charging pump OPERABLE, restore at least two charging pumps to -OPERABLE status within 72 hours or be is at least HOT STANDBY and borated to SHUTDOWN MARGIN equivalent to at least L1. ) Ak/k at 200*F within the next 6 hours; restore at least two charging pumps to OPERA 8LE status within the next l 7 days or be in COLD SHUTDOW within the next 30 hours. SURVEILLANCE REQUIREMENTS 4.1.2.4.1 The required centrifugal charging pump (s) shall be demon =trated OPERABLE by' testing pursuant to' Specification 4.0.5. 4.1.2.4.2 The required positive displacement charging pump shall be demonstrated OPERABLE by testing pursuant to Specification 4.1.2.2.c. g 4.1.2.4.3 Whenever the temperature of one or more of-the Reactor Coolant System (RCS) cold legs is less-than or equal to 350*F, a maximum of two charging pumps shall be OPERABLE, except when Specification 3.4.8.3 is not applicable'. When required, one charging pump shall be demonstrated inoperable # at least once per 31 days by verifying that the motor circuit breakers are secured in the open position. "The provisions of Specifications 3.0.4 and 4.0.4 are not applicable for entry into MODES 3 and 4 for the charging puen declared inoperable pursuant to - Specification 3.1.2.4 provided the ch ing pump is restored te OPERABLE status within 4 hours efter entering MODE 3 or prior to the temperature of one or more of the RCS cold legs exceeding 375'F, whichever comes first.

- In M00E'4 the positive displacement pump may be used in lieu of one of the required centrifugal charging pumps.

An inoperable pump may be energized for testing provided the discharge of the pump has been isolated frcm the RCS by a closed isolation valve (s) wit $ power removed from the valve operator (s) or by a manual isolation valve (s; secured in the closed position. l l COMANCHE PEAK - UNIT 1 3/4 1-10 Amendment No. 5 l

t . s > -4 '$$ A' N' -. "u z rv re l Alll l II4 o ~c u t RAlll Allori Murill0Ritit. ItJ'.lRIIMi tal Allori 10R l>l ArJI g)l I gAllort, - ru Ti ~ C. ,i., Mita lMilt1 l' filAlifili'. tilAfifil I ', Al'I'l It Allii Al ARM / IKil' l ist.iC l 10..l.lAl. titil l 10 IRil'/Al ARM Ol'l RAlll i Multi *. _' i ll'O ltil AfIlOf4 I. ItC', i e.ek.up 'lletec t ion

~'

l' art is ul,.ste Radioat t ivi t y ti A. I

1. ?, l. 4 f4 A.

/# a. le. Gascasen, Ratlicact ivity fL A. I 1 ?, 3, 4 14 A /** i ?. Einitainment Vent ilat ioni I ses !.it ion G,iscous R.selio.it_ t iv i t y I I I, /, I, // i 4,t.** 3. Cusil ol lloom m A i r-laitake*-R.ielial insi I e vel I / i n t.it e-7/ i sit,it i-AlI .h4-10~'l?il pt. t /m I .o r I i i

TXX-92001 ATTACHMENT 27 PAGE O 0F 41

E aC7:e :::..'.' ;. E Eu

-C'

'ANCB.v.

5.'.'CI L'. ANCE REQUI REMENTS ( Conti nvec ) 44'.2.2

- e recut e: steam generators shall be determined OPERABLE By[grifyirQ secondary side water level to be greater than er exal a.

tolLJ.(narrow range) at least once per 12 hours, and By performing the surveillances pursuant to Specification 4.0.6. c. 4.4.'.2.3 ine reovired reactor coolant loops shall be verified in operation anc c rculating reactor coolant at least once per 12 hours. to m

w- - fff20 2T REAC'OR ' COOL ANV 3'5TE4 -PAGE-9. 0F 41 H07 SsuTOOWN L. LIMITING CONDITION FOR OPERATION ~ .3.4.1.3 At least two of-the loops listed belo shall be OPERABLE and at least one of these loops.snall be in operation:" Reactor Coolant _ Loop 1 and its associated steam generator and 4. reactor coolant pump,** Reactor Coolant Loop 2 and its associated steam generator and b.. reactor coolant pump,** Reactor Coolant Loop 3 and its associated steam generator and c. reactor coolant. pump,** Reactor Coolaut ~ Loop 4 and its associated steam generator and o, reactor coolant pump,** e. RHR Loop =A,-or f, 'RHR Loop B. I 'APPLICAB!LITY: MODE 4 ACT!0N: 1 'aitnlless than the:above recuired Icops OPER-3LE, immediately a.-- initiate corrective action to return the_ required loops to'0PERABLE status as soon as possiblei if.the remaining OCERABLE loop isfan RHR loop, be in COLD SHUTDOWN within-24-hours. l L i - (

All' reactor coolant pumps and-RHR pumps may be deenergited for up to 1 nour-U provided:

(1)~oo operations are permitted that would cause dilution of tne Reactor Coolant System boron concentration,-and (2) core outlet temperature is maintained at least 10*F below: saturation temperature. "*A reactor coolant pump-shall not :be started in Mode a unless.the-secondary [ watertemperatureofeachsteamgeneratorislesstnan[50fFaboveeacnof ? the Reactor Coolant System cold leg temperatures. 1 r' COMANCHE FEAK - UNIT 1 3/4 4-4 i ( 4 _.--.-.,,.7 m - _-,.~,._,,.m.m_

TXX.92001 ATTACMMENT 27 F AGE 10 0F 4j;;c.p --.. s. 9,3 gy l -Of 5 HUT 00WN .!MIT:NG COND:'!CN COR OPERATION AC'::N: i::-t sec) i ditn no loop in operation, suspena all operations involving a b. recuttien in boron concentration of the Reactor Coolant System and 'immediately initiate cor ective action to return the required 1000 to operation. SURVE!LLANCE REQUIREMENTS 4.4.1,3.1 The required reacter coolant pump (s), and/or RHR pump (s) if not in operation, shall be cetermineo OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability, 4.4,1.3.2 The required steam generator (s) shall be determined f/ERABLE By[1.(narrowifying secondary side water level to be greater than o a. e to range) at least once per 12 hours, anc By :erforming the surveillances pursuant to.:ecification 4.0.6. c.

4. 4.1. 3. 3 At least one reactor coolant or RHR loop s* 311 be verified in operation and ci-culating reactor coolant at least one-cer 12 nours.

C;MANC-E :E M - cN!T 1 3/4 4-5

TX X-92001 ATTACHMENT 27-PAGE110F..$l..,............g

e

2 :

LO 5-UTDOWN * ;3005 :!L.E0

.. u. - t. yg

r. yo t. *. *. y : n. q r. e e. : s * *. y.

..86est.4-

3. 4 1.4.1 at ' east One resicua!

eat retovni (RHR) ':c; s a

eC E:aB;E s :

::e st':**, 3: e' t er: a.

re a: 'ti;nal RnR 1:00 5 al' Oe ;PERAELi'*, :r b.

The see:ncary si e aster levei ;f pt least t.c steam ge e#5t;+s 0 sn4 11 te greater tran 0* ecual ta[lhi(nar0 range). ADOL!CaBILITY: MODE 5 *ith eeactor ::oiant ';;;s fiec'** z AC'!CN: 3.

ita One of tNe ELR icoos iaccerable or with less than t e recu' te steam generator water level, immeoiately initiate c0rrective act';n to return the inoceracle RnR iceo to OPERAELE status or rest:re e

re:uirec steam generator aater level as soon as possible. c. With no RHR 1000 ir coeration, sus end all Operations-inv:1.irg a recuction in Dor 0n : ncentration of the Reactor Coolant ijste?. anc immeciately initiate correttive action to return the recuirea RrR loop to operation. SURVE:LLANCE REQUIREMENTS

2. 4,1. 2.1.1 Tre seccrdary si e ater. level of at lea:-

d t-o steam ;ecerat:rs aren recuirec snall ce cetermi6ec to ce within limits 5*. least once :ec 12 hours.

4. 4.1. 4.1. 2 At least one RbR loop shall be ceterminec to te in operati n anc riaculating *eactor coolart'at least once per '2 hours, 4

The RHR pump may be deerergizec for up to I hour provicec:

(1) no c:eraticns are permitted that would cause dilution of-the Reactor Coolant 5ystem coron concentration, and (2) core outlet temperature is maintainec at least IC'F below saturation temperature.

- 0ne RHR 1000 may te inocersele for up to 2 hours for surveillance testing; provicea tne otner RHR loco is OPERABLE and in operation.
  - A reactor coolant pumo snall not be startec in Moce 5 unless the seconcary water temoerature of each steam generator i less than 5'0*F 30cve eacn of tne Reactor Coolant System cold leg temoeratures.

COMANCHE DEAK - UNIT 1 3J4 4-6

o 2o a $IIU "' M 1, -xm N3O- '. l I. A._lli t_ _4_. _4_- /. - 5:5 ~ 3* y, _um Hi AI.lijl! %i Mil d. MA11 RlAl,',IIRVI,111 ANCl, I'R0gl(AH - W1 till)RAWAl_ _,(fit litilI y IB cal">lli i VL 55l L tillHill1: 10 cal 10N _ li All i At:luit WIliiDNAWAl iIMI 11 Sil. !s"

4. t:0 ist Heitn linig Y

711.O (f.),8 I l'Y l. f ') V fi l. u"

1. f.')

'J l i l'Y x ?lh.5" 4.00 IS Ill*Y W 4.00 ' t.inelley 4se s. / .101. 5" 4. 1181 'a S t.indley Y

TXX-92001= ATTACHMENT 27 PAGE 13 0F-41 l i l l l l l l l l } l l I 7 '..~l l l l l l l l l l l l I I '" l I I I i i I I I i 1 I I i i l l l l I I I M I i. l l l I I I I I I IN, i j l l l l l l l l l l I I I i l l l l l l l I I I i z s s '. [ l l l j l l l l I I I s l I I I I l i I I I I I I i l l-1 I I I I I I i l i I a.. ~ i I ) l l l l l l ,1 l l i ~" l l { } } l l I I l l I I i i ~" I l l l l l l l l l l l ~ } l { I l l l l l l l I l l l l l l l -l l l l l i s:: 60 100 540 180 200 20 300 2'I RCS TEMPER ATURE. OEG F Ih!GURE3.4-4 PORV SETPOINTS FOR OVERPRESSURE MITIGATION L APPLICABLE UP TO 15 EFPY C."MANC-E ; Eld.-, UNIT 1 3/4 d*29

TXX-92001 ATTACHMENT 07 PAGE 14 0F 41 iu!4GENC# ::RE C:h:N3 i 5'Eu; 3/4.5.1, ACCUMULATORS C%D tiG INJECTION LIMIT;NG CONDIT!ON CCR ODERATION 3.5.1 Eac. co;; 'e; jection accumulator shall :e CPERABLE witN The cisc9arge isolation valve open with power removed. a. Anincicatedtorated.aterlevelofbetween[3k.and[6h, c. A boron concentration of between[1900]and[2200] ppm, and c. An inoicated cover-cressure of between 623 ano 644 osig. c. ADDLICAE!LITY: MOCE5 1, L and 3* ACTION: With o e co'c 'eg i,jection accumulator inoperaole, except as a result a. of a closec isolation valve or the boron concentration outside the requirea alues, restore the inoperaole accumulator to OPERABLE status within 1 nour or be in at least HOT STANDBY within the next 6 nours anc recute cressurizer pressure to less than 1000 psig witnin the folle.ing 6 hours. Witn one cc!c leg injection accumulator in; D.

able cue to the isolatien-,a;ve Deing closed, either immec' ely open :ne isolation salve or te in at least HOT STANDBY within 90urs and reduce Dress.ri:er cre55ure to less than 1000 psig.ithin the following 5 nours.

Witn the :or:q concentration of one coic leg injection accumulator c. outsice 19e recuireo limit, restore the Dort c:ncentration to within the *ecu ce: limits within 72 hours or be i* at ' east HCT 37ANCBY i aithin t e next 6 hours and reduce pressurice* oressure tc ?ess than 100C Osig within the following 6 hours, $URVEILLANCE RE0010EPENTS a.5.1.1 Each cold 'eg injection accumulator shall be cemonstratec CPERABLE: at least : ce ter 12 ' curs by: a, 1) Ver* *.ing tne indicated Dorated water ie.ei anc nitrogen cover cressure in the tanks, and

Dressari:er press.:e ace e 1000 csig.

CMANCHE OEAO - l,lN!T ; 3'45-1

1TXX-92001 ATTACHMENT 27 -PAGE 15 0F 41 EMERGEN:':CRECOOLING5'STEMSL 345.4 REFUELING WATER STORAGE TANK LIMITING CONDITION FOR OPERATION 3.5.4 ine refueling water storage tank (RWST) shall be OPERABLE with; Aminimumindicatedboratedwaterlevelof[95h, a. A boron concentration of between 2000 and 2200 ppm of boren, D. A minimum solution temperature of 40'F, and c. d. A maximum solution temperature of 120'F. APDLICABILITY: MODES 1, 2, 3, and 4 ACTION: With the RWST inocerable, restore the tank to OPERA 6LE status within 1 hour or be in at least HOT STANDBY witnin 6 nours anc in COLD SHUTDOW following 30 hours. SURVE!LLANCE REQUIRCMENTS 4.5.4 The RWST shall be demonstratec OPERABLE: At least once per 7 days by: a. 1) Verifying the indicated berated water Iv e1 in the tank, and 2) Verifying the boron concentration of the water. b. At least once per 24 hours by verifying the RWST temperature when the outside air temperature is less than 40 F or greater than 120 F. COMANCHE DEAK ' UNIT 1 3/4 5-10

m TXX.92001- - ATTACHMENT 27 - PAGE;16 0F 41 y 3 'l *. -; ; N * ; Sa..g.,. ;, 3 g. ; 3'A* l 031 MAD- ::NTA!NMEN' i N'::NMENT :N'!0a 'v ..y...,..... 4 u...,......,4 a. M, ..gm.....y ..a .w 3. *. ;,.' - De d ma ;. :;Nia!', MIST *NTE3RITr snall :e maintat.ec. s 2 3,..,Bi.. w:. ...a ...;) ...;, anc *. ACT! N: Wit 9out primary ::NTAINMENT INTEGRITY, restore CCNTAINMENT INTE;RITV s in t a. 1.nour or ce in at least HOT STAN08Y.ithin the. next 6 nours anc in C;LD Shut;0WN wstnin :ne followfog 30 hours. IV:VE*LLANCE_ RECu:9EMEN'i

4. 6.1.1 : a i m a r;. 0NTA,'W'EN'

NT EGR;'t snall te dem0nstratec:

At-least once per 31 cays ey verifying that all cenetrat' ~s" act a. capacle of being closec Oy OPERABLE contair t automatic is: Int :n valves and required to be closed curing acc.ent conditions are closec cy valves. Olind. flanges, or ceactivt to aut:matic valves securec in'treir-:ositions, except as provi:-: in Table 2.*..i of tre Tecnn':al Recuirements Manual. $;. <eri'yiag trat each containment air lock is in c:moliance wi*n O.

ne -ec.irements of Scecificatic., 3 J.,1.2;- arc L

Af ter each c.icsing of each penetration-subject to. Type B testing, c. -excect tne containment air locks, if opened following a Tjoe A ce B test. :) leak rate testing the seal with gas at a pressure not less inan P,[i8.'y-asig, and verifying that when :ne measured leakage rate L 'ertMlsesealsisaccedtotheleakageratescetermineapursuantto Specification 4.6.1.2d. for all other Type B anc C penetrations, L the combined leakage rate-is less than 0.60 L - a '"Except_ <al<es, clinc f:anges, and ceactivated autornat'ic valves which are lccatac i*sice the containment and are locked, sealed or otherwise securea in the closeo position. These penetrations shall be-verified closeo curing eacn COLD SHUTOCwN except tnat such verification need not te perforcea more often than once cer 92 days. The blind flange on the fLel transfer canal neec not be ve-ifiec c10sec eAcept af ter each drainage of tne canal. CCMANCHE :EAK - W ' ' P c-L

IXK 92001 ATTACHMEWT 27 **N?i!w!NT !.p(N PAGE 17 0F 41 CONta!wrNT LtaxaCE L:ul?!NG10N0!T!09FCRcotoAT:CN l (

3. 6.1. 2 Cont s'ntent leakage rates shall te litit ted to:

g a-an overs 11 integratec leakage rate of: 1) Lest tnan or equal to L,, 0.10% by weight of the containment M airper24hoursatP,,[48.3]psig,or [*h 2) Less than or ecual to L, 0.05\\ by weignt of the containment g air per 24 hours at a reducea pressure of P, 24.15 psig. g b. A combined leakage rate of less than 0.60 L, fer all penetrat*,ons and ',alves subject to Type B and C tests, when pressurized to P,. APP.LICABILITY: MODES 1, 2, 3, and 4 D ACTION: With either the meas'ared overall integrated containmeat leakage rate enceecing 0.75 L, ca 0.75 L, as applicaole, or the measured ec ined leakage rate for g all penetrations And valves subject to Types B and C sts exceeding 0.60 L,, renore the overall integratec leakage rate to less t n 0.75 L, or less than O 75 L, as applicable, and the combined leakage ra'te sr all penetrations g 4 subject to Type B and C tests to less than 0.60 L prior to increasing the Reactor Coolant System temperature above 200'F. a e LURVE!LLANCEREQUIREMENTS 4.6.1.2 The containment leakage rates shall be demonstrated at the following test schedu19 and shall be determined in'conformance witn the criteria speti-5 fied in Appendix J of 10 CFR 50: 4 Three Type A tests (Overall Integrated Containment Leakage Rate) a. shall be conducted at 40 : 10 month intt!<als during shutoown at apressurenotlesstnaneitherP,,[48.3Jpsig,oratP.24.15 g psig, duri q -sch 10 year service period. The third test of each set shall be corducted during the shutdown fer tne 10 year plant inservice inspection: C^MANCHE PEAK - UNIT 1 3/4 6-2 . - - - - - - - - - _. - - - - - - - ~ - - - - - - - - - ' - ' - - - - - - - - - - - - - - - - ' ' - - - - -

T X X-92001 ATTACHPEM7 27 PAGE 18 0F 41 __N_' a ' '.u f S " !.;*{ug

NTA:suENT AIR LOCK 5

!*

NG CONDITION r;n r,*:r:: ::N 1 t. '. 3 Ea:

-ta*- ent aic 'ock snail ce OPE 8ABLE.it-a.

Sotn ::ces :'osec e, cept.ren tre sic 'ocs's:ei; sec*:r : +s-traesit ente f 1*0 edit tar;sgn tae

ntainment, t,m e n 3t ',3g; :.9 sir 1::( ::or s'ai' e :':sec. Inc o.

An overall air 1 0: ( 1eakage rate of less inan or ecua! :: : ;5 L at ?,.@B.yostg. 5 APOL::AL: LI*Y: M00E5 1. 2. 3. ano 4. ACT!ON: a. With ore c0ntainment air 'oCK coor inc;erable; 1. Maintain at least t*e OPERABLE air lock ccer close a a eit er restore tae incoeracle air lock coor to 0FERABLE stat.s.itnin 24 hours or lock tne OPERABLE air lock coor : losec. 2. Operation maj then continue until per*- ance f t*e e.- recuirec overall air lock leanage test oviced that tne ::E:42Li aie lock coor 's verified t be locke: osec at least Once :ea 31 cays: 3. Ot'erwise. :e in at least HOT STANDBY tnin tne aeit e curs and in COLD SHUTDOWN within the followia; 30 notes; ie 4 Tne provisions of 5pecification 3.0.4 a e not acci>:2 e. c. with the containment air lock inocerable. ta:ect as tre es. 't :' an inocereole air lock door, maintad n at least one air locs cecr closec; restore the inoperable air lock to OPERABLE status within 24 tours or be in at least HOT STANDBY within the ncxt 6 hours anc in COLD SHUTOOWN within the following 30 hours. i COMANCHE PEAN pHIT 1 3/4 6-4

TXX. 92001 ATTACH"[fli 27 PAGE 19 Of 41

4?A! Nuts' !.!*tus Saa t!LLANCE REQUlGEMENTS v

J. 5. ' 3 !):h C'*tA'** eat aie '0:4 InaII te Je90nstrgte: CDE2A$t!: 1. a ' '.

4 * *2 "Ourn f011 0a*Ag eaCn O'Ollag, eacect. hen t*e

' ;:n *l tetr@ sies f0r mul 1 t'0 e entries. t*en at least Once gee 72 a;'.rs, *j .ertfy'ay leal leakage 'l ' ell *'In 0,01 L ll :ettemtrec :,, prec'gita '1 0 ?ttlwretents. hen *eangred for at lea lt 30 Se;0act.its the /Olute tetat t*e 5elli at i OCnttent ;ressure Of

'etter than or eQwSI10[,46, 015 G '.

D. If Conducting overall Sir 10Ck leakage te$ts at not legg t *, a n D (jB.$f elig, and verifying the overall air iock leakage rate is within its limit; '. ) at least once per 6 mentns,' and 2) Dedor to estaelisning CONTAINMENT INTEGRITY -hen maintenance te-'or+ec en tre air lock that cou!c affect tre air as :eea loca sealia; :s:aci14ty. at least Once ce" 5 tentns by serifying tnat only one c::r c +n ea n li" '0:k Oln be Ope *ed at a time.

The pro'.-isions of $cecification 4.0.2 are not applicable.
- This re:resenti in e= emotion to 10 CFR 50 Appendix J, paragrach }lI.O.2(b)(ii).

COMANCHE :!AK UNIT 1 3/4 6 5

TXX-92001 Af?ACHMENT 27 PAGE 20 0F 41

'catw!N' i<st!us

_N'EGNAL 2RE550RE ,:utt:NG CON 0! TION r0R OPE 0ATION

3. 6.1. 4 De Mary contd nment internal inoicatea pressure sna11 ee maintaineo cet een[0.330sig and 1,3Josig.

i:Dt!CA9: LITf: MODES 1. 2, 3, anc 4. 40* 04: I .stn the containment internal pressure outside of the limits above, restore tre internal pressure to within the limits within 8 hours or be in St lesst *0T i iTANCBy.ithin the next 6 hours ano in COLD SHVT00wN witnin the 'oilo.ieg 30 4 acurs. .i 50R', E :LLaCE REQUIREMENTS ' i.1.4 're primary containment internal pressure sna : te cetero e:.: ce it m t e Umits at least once eer 12 hours. I 1

MANCHE EAA

UNIT 1 3 4 6-6

-__-*-,y-y- ' - - ' - - *- * * - ~ '

TXX-92001 ATTACHMENT 27 i OAGE 21 0F 41

s a p,ur.,*

p4;

c truef:4'LRE s:u:T N CONC:':CN rCR C85Ga* N

3. 5.1. 5 n, :: ta 9 rent s'.erage air te rea:ture shall not en:ee:{'; }c.

m.::ss:u.: e ts 1. :. 3. ane 4 AC '!:-N: uththecentainmentaverageairtemperaturegreaterthan[120)F,recutetre average ai temocrature to within the limit within 8 hours, or te in at least HOT STANDBY.ithin the nest 6 hours ano in COLD SHUTDOWN within the following 30 hours. SURVE: LLANCE REOUIREMENTS 4 ,1. 5 me crimarj containment aversge air temperat. e Shall te tre lojulted aserage o' t.o temperature $ at or ato,e tne following ::ntainment locatd nl Cf .hi:h at 'et$t :9e tem 0erature is from Ic:stion s. or a:uve and snall te cetermined at least once per 24 hours:

t iti oj tere. E'. [ic00 -6'3 a.

Floor, E1.{a60'-0"3 C: MAN;r! ;ik. gNlT 1 3,4 6 7

TXX-92001 ATTACHMENT 27 PAGE 22 0F 41 : LAN

.;- u;

NCENSA*! STORACE iaNA r!91 TING l0N0!*:0N C**

ODERATION 3.'.;.3

e :

e inte rage tar ( (CST) shall be OPERABLE.ith sn incicatec

.ater 'ecel O f i t

e t t '.

ADDLIAEI.liv. "CCE3.', 2, and 3. AC':0N: With the CIT incoerable, within 4 hours either: Restore tre CST to OPERABLE status or be in at least HOT a. STANC2' witnin the next 6 hours end in HOT SHUTDOWN.ithin the fo11cwing 6 hours, or b. emonstrate t e OPERABILITY of the Station Service kater (SSW) s < stem as a :ackuo sutoly to the auxiliary feed.ater cumes anc estere tne 0.*T to OPERABLE status within 7 days or be in at least ~0T S*ANDBf within the next 6 hours and in HOT SHUTOOWN within the 'oilowing 6 nours. 3LRtE:'..ANCE Ai' t:4!ugs 5 4. '. L 3, '. The CS' s a': ce cemonstrated OPERABLE at +ist once eer 12 neurs oy er*f 'rg ine in:4:stea.ater level is within its limits when the tant is the su o' 5:ur:e 8 0 t e awxiliary feedwater pumps. 4.L*.3.2

- e ila i 6:41 shall be demonstrated OPERABLE at least :nce per 12 Mcurs rene.er t*e 55W system is being used as an alternate sucoly scur:e t:

the auxili try retc.ater pumps by verifying the SSW system OPERABLE and ea:n motor cotentec.a!,e :et een the SSW system and each OPERABLE auxiliary 'eec- . ate" o#to is OPERAELE.

MAN:.wE ;!As cNIT 1 3/4 7-5

1 TXX 92001

'N' I'5TEuS IAffACHMENT 27

..!!!r! ACT!v!TV l PACE 23 0F 41 LIMIT!NG CONDIT!0N FOR OPERATICN

3. 7.1. 4 Ihe sDeCifi$ activity of the SeCorcery Coolant System shall be less than or equal to (0.1JmicroCurteigram OOSE ECU! VALENT l 131.

APPL!cA9!L!T!: MOCES 1, 2, 3 anc 4 ACTION: WiththespecificactivityoftheSecondaryCoolantSystemgreiterthan(0.1) microcurie / gram DOSE E0V! VALENT I-131, ce in at least HOT $TAND6f witnin 6 hours and in COLD SHUTDOWN within the 'ollowing 30 aours. SURVE!LLANCE RE0V!REMENTS 4.7.1.4 The specific activity of the Secordary Cot la-' System sna11 te cetermined to be within the limit by performance of t - sampling and analysis program of Table 4.7 1. CDMANCHE PEAK - UNIT 1 3/4 7-5 w

- - - - ~.. _ TXX.92001 etaNT s,f'Eus ATTACHMENT 27 PAGE 24 0F 41 3,4. 7. g ontwAny otANT VENT!LATION sr$7EM ESF FILTRAT!0N UNITS L!MIT!NG CON 0! TION FOR OpfRAT!CN 3.7.8 Tao iscepencent ESF Filtration Trains sna11 be OPERABLE. ADDLICABILITe MOCES 1, 2, 3. ano 4 ACTION: With one ESF Filtration Trsin inoperable, restore the inoperable ESF a. Filtration Train to OPERABLE status within 7 days or be in at least HOT STAN0BY within the next 6 hours and in COLD SHUTDOW following 30 hours, b. With the Inability to reach and maintain a negative pressure in the negative pressure envelope of the Auxili -Buildings greater than or equal to(0,05] ary, Safeguards, and Fuel inch'bater gauge, restore the PRIMARY PLANT VENTILATION SYSTEM to OPERABLE status within 30 da be in at least HOT STANDBY within the next 6 hours and in COL within the following 30 hours. With the inability to reach and maintain a negative pressure in the C. negative pressure envelope of the Auxiliary, iafeguards, and Fuel Buildings greater than or equal to 0.01 inc..ater gauge, restore the PRIMARY PLANT VENTILATION SYSTEM'S ability t: maintain a negative pressure of greater than or equal to 0.01 te;n water gauge wit *in7 days or be in at least HOT STAN0BY within tP4 next 6 hours and in ' COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.7.3 Eacn ESF Filtration Train shall be demonstrated CPERABLE: At least once per 31 days on a STAGGERED TEST BASIS by initiating, a. from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that each ESF Filtration Train operates for at least 10 continuous hours with the heaters operating; At least once per 18 months or (1) after any structural maintenance b. on the HEPA filter or charcoal adsorber housings, or (2) followin painting, fire, or chemical release in any ventilation rome com g municating with the system by: 1) Verifying that each ESF Filtration Unit satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 1.0% by using the test procedure guidance in Regula-tory Positions C.5.a. C.S.c, and C.S.d of Regulatory Guide 1.52. COMANCHE DEAK - UNIT 1 3/4 7-20 \\ ... _ =

TXX 92001 ATTACHMENT 27 PAGE 25 OT 41 ?a! !.!:*4:04 ChE# SYSTEMS l 'a e.1 a.C. $00RCES CDEDAT!NG . wlt:NC ::ND1? 0N FOR CD!aATION L S.1.1 es a minimum, the following A.C. electrical power sources shall ce ODERaBLE: Two physically iroependent circuits between the offsite transmission a. network ano the onsite Class 1E Distribution System, and c, Two separate and independent diesel generators, each with: 1) A teoarate day fuel tank containing a minimum volume of[1440) i gallons of fuel, 2) A separate Fuel Storage System containing a minimum volume of [B6,000] gallons of fuel, and 3) A separate fuel transfer pump. ADDLICABILITY: MODES 1, 2, 3, and 4. ACTICN: With one offsite circuit of the above*reouired A.C. electrical oower a. sources inoperable, cemonstrate the OPERABILITY of the remaining A.C. Sources by performing Surveillance Requirement 4.8.1.1 la.ithin 1 nour and at least once per 8 hours thereafter. If either diesel generator has not been successfully tested v' thin the past 24 hours, oemonstrate its 0?tRAB!LITY by performing $, ve111ance Requirement 4.S.1.1.2a.4) for each such diesel generator separately, within 24 hours.' Restore the offsite circuit to :: ERABLE status within 72 hours or be in at least HOT STAND 8Y within the next 6 hours ano in COLD SHUTOOWN within the following 30 hours. b.

ith eitner diesel generator inoperable, demonstrate the OPE 4ABIL;TY of the acove required A.C. offsite sources by performing Surveillance Requirement 4.8.1.1,14, within 1 hour and at least once per 6 rours thereafter.

If the diesel generator became inoperable due to any cause other than preplanned preventive maintenance or testing, oemon-strate the OPERA 81LITY of the remaining CPERABLE diesel generator by performing Surveillance Requirement 4.8.1.1.2a.4) within 24 hours unless the diesel is already operating and loaded. Restore the inoperable diesel generator to OPERABLE status within 72 hours or ce in at least HOT STANOBY within the next 6 hours and in COLO $ HUT 00aN within tne following 30 hours.

During cerformance of surveillance activities as a requirement for ACTION statements, the air roll test shall not be performed.

COMANCHE :!AK UNIT 1 3/4 B 1 -~ ~

TXX 92001 ATTACliMENT 27 PAGE 26 0F 41 EL!:*R!Ca. :: ER l'5?iPS a C. SOURCES SWUTCCwN LIMIT!NG CONDITION FOR OPERATION

3. 9.1. 2 as a minimum, the following A.C. electrical power sources snail ce

.0PERABLE: One circuit between the offsite transmission network and the Onsite a. Class 1E Districution $ystem, and b. One diesel generator with: [ i 1) Day fuel tank containing a minimum volume of[1440 gallons of futi. 2) Afuelstoragesystemcontainingaminimumvolumeofh6.000] gallons of fuel, and 3) A fuel transfer pump. APDL I C AE !'. ! T v : MODES $ and 6. i AC'!CN: 'With less than the above minimum required A.C. electr.s1 power sources OPERABLE. immeciately suspend all operations involvin; *0RE ALTERATIONS, positive reactivity _ changes, movement of-irradiated fL41, or crane coeri. tion with loacs over the fuel storage pool, and within 8 he.rs, depressurize ano vent the Reactor Cociant System-through a greater than or soual to 2.98 savare inen vent. In aceition, when in MODE 5..'*h the reactor coolant locos not filleo, or in MODE 6 with tne water level less than 23 feet above tne reactor vessel flange. immediately initiate corrective action te restore the reouired sources.to OPERABLE status as soon as possible. SURVEILLANCE REQUIREMENTS

4. 8.1. 2 The above required A.C. electrical power sources shall be comonstrated OPERABLE by the performance'of each of the requirements of Specifications 4.2.1.1.1. 4,3.1.1.2 (escept for Specification 4.8.1.1.2a.5)), ano 4.8.1.1.3.

1 i COMANCHE PEAK # UNIT 1 3/4 B-10 i

TXX 92001 ATTACHMENT 27 PAGE 27 0F 41 SDECIAL *!$7 EtCEDTICN$ 3/4.10.a REACTOR COOLANT LOOPS LIMIT!NG CON 0! TION FOR OPERATION i 3.10.4 The-limitationa of Specification 3.4.1.2 may De suspended during the I performance of not roc drop time measurements in MODE 3 provided at least two reactor coolant loops as listed in Specification 3.4.1.2 are OPERABLE. A9DLICABILITY: During performance of het rod drop time measurements. ACTION: With less than the above required reactor coolant loops OPERABLE during the performance of hot rod drop time measurements, immediately open the reactor trip breakers and comply with the provision of the ACTION statements of f Specification 3.4.1.-2. 2 i SURVE!LLANCE REQUIREMENTS 4.10.4 At least the above required reactor coolant

es shall be determinec OPERABLE within'4 hours prior to the initiation of he red drop time measure-ments by verifying correct breaker plignments and ine :sted power-ave' ; 1 ability

-and by veri g the indicated secondary side water '..el to be gre; er than or equal to narrow range. 4 'I ~ COMANCHE' PEAK - UNIT 1 3/4 10-4 t 4+-=.- -w. -<-n>=erere-+>% ew w, .mo,, re r e w m. nu -, -err--e rr-*--*- ec-o w

-r-w+---*-e,u-w-e-

-,v,3.r-,,,rwe ye=w r, -,,-e-m+-we-rs+-,=----e-vw--s- -rm.-----erw-,-w3-ve'- -w'

TXX-92001 ATTACHM:NT 27 PAGE 28 0F 41 3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 BORATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN enstres that: (1) the reactor can be made subcritical from all operating conditions, (2) the reactivity transients associated with postulated accident conditionr, are controllable within a ceptable limits, and (3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition. SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS T,yg.The most restrictive condition occurs at EOL, with T,yg at no loading operating temperature, and is associated with a postulated steam line break accident and resulting uncon-trolled RC MARGIN ofi)l.gE ak/k is required to control the reactivity transient.In th (ooldown. Accordingly, the SHUTDOWN KVtGIN requirement is baseo upon this limiting condition and is consistent with FSAR safety analysis assumptions. With T,yg andisbasedontheresultsofthebor(ondilutionaccidentanalysis.less Since the actual overall core reactivity balance comparison required by 4.1.1.1.2 cannot be performed until af ter criticality is attained, this comparison is not required (and the provisions of Specification 4.0.4 are not applicable) for entry into any Operational Mode within the first 31 EFP0 following initial fuel load or refueling. 3/4.1.1.3 MODERATOR TEMPERATURE COEFFICIENT The limitations on moderator temperature coefficient (MTC) are provided to ensure that the value of this coefficient remains within the limiting condition assumed in the FSAR accident and transient analyses. The MTC values of this specification are applicable to a specific set of plant conditions; accordingly, verification of MTC values at conditions other than those explicitly stated will require extrapolation to thc.e conditions-in order to permit an accurate comparison. The most negative MTC value equivalent to the most positive moderator i density coefficient (MDC) was obtained by incrementally correcting the HDC j und in the FSAR analyses to nominal cperating conditions. These corrections l l COMANCHE PEAK - UNIT 1 B 3/4 1-1 Amencment No. 5 ~.

ThX-92001 ATTACHMENT 27 PAGE 29 0F 41 REACTIVITY CONTROL SYSTEMS BASES MODERATOR TEMPERATURE COEFFICIENT (Continued) involved subtracting the incremental change in the MDC associated with a core condition of all rods inserted (most positive M)C) to an all rods withdrawn condition and, a conversion for the rate of change of moderator density with temperature at RATED THERMAL POWER conditions, This value of the MOC was then transformed into the limiting End of Cycle Life (EOL) NTC value. The 300 ppm surveillance limit MTC value represents a conservative value (with corrections for burnup and soluble boron) at a core condition of 300 ppa equilibrium boron concentration and is obtained by making these corrections to the limiting EOL MTC value. 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 remains within its limits since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup. 3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the r@ actor will not be made critical with the Reactor Coolant System average temperature less than 551'F, This limitation is required to ensure: (1) the moderator temperature coefficient is within it analyzed temperature range, (2) the trip instrumentation is within its normal operating range, (2) the pressurizer is capable of being in an OPERABLE status with a steam bubble, and (4) the reactor vessel is above its minimum RTHOT temperature. 3/4.1.2 BORATION SYSTEMS TheBoronInjectionSystemensuresthatnegativereactivitycontrolis available during each mode of facility operation. The components required to perform this function include: (1) borated water sources (2)chargingpumps (3) separate flow patha., (4) boric acid transfer pumps, an,d (5) an emergency, power supply from OPtedBLE diesel generators. With the RCS average temperature above 200'F, a minimum of two boron injection flow paths are required to ensure single functional capabilit the event an assumed failure renders one of the flow paths inoperable. y in The boration capability of either flow path is syfficient to provide a SHUT 00WN MARGIN from expected operating conditions of L1.6A Ak/k after xenon decay and cooldown to 200*F. The maximum expected boration capability requirement occurs at EOL from full power equilibrium xenon conditions and requires Q5,700lgallons of 7000 ppa borated water from the boric acid storage tanks or U0,7027ga11ons of 2000 ppm borated water from the refueling water storage tank (RWST). COMANCHE PEAK - UNIT 1 8 3/4 1-2 Amendment No. 6 I

TXX-92001 I ATTACHMENT 27 PAGE 30 Of 41 REACTIVITY CONTROL SYSTEMS BASE 5 BORAT10N $YSTEMS (Cont.inued) With the RCS temperature below 200'F, one Boron injection System is acceptable without sing 16 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 maximum of two charging pumps to be OPERABLE and ;he requirement to verify one charging pump to be inoperable below 350'F provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV. The limitation for minimum solution temperature of the borated water sources are suf ficient to itraent boric acid crystallization with the highest allowable boron concentration. The boron capabi ity requirsd below 200*F is sufficient to provide a SHUTDOWH KARGIN of[1. Ak/k af ter xenon decay and cooldown from 200*F to 140*F. This condition requires eith boric acid storage tanks or[er[1,]100lgallons of 7000 ppe borated water from 7,113 ga11ons of 2000 ppe borated water from the RWST. As listed below, the required indicated levels for the boric acid storage tanks and the RWST include allowances for required / analytical volume, unusable volume, measurement uncertainties (which include instrument error and tank tolerances, as applicable), system configuration requirements, and other required volume. Tank HODES Ind. Unusable Required Measurement System Other level Volume Volume Uncertaint) Config. (gal) (gal) (gal) (gal) RwST 5,6 24% 45,494 7,113 4% of span 57,857 N/A 1,2,3,4 95% 45,494 70,702 4% of span N/A 357,M5al \\ Boric 5,6 10% 3,221 1,100 6% of spar N/A N/A Acid 5,6 20% 3.221 1,100 6% of span 3,679 F/A i Storage (gravity feed) Tank 1,2,3,4 50% 3,221 15,700 6% of span N/A N/A J The OPERABILITY of one Boron Injection System during REFUELING ensures that this system is available for reactivity control while in H00E 6. EAdditional volume required to meet Specification 3.5.4. COMANCHE PEAK - UNIT 1 B 3/4 1-3 Amendment No. 5

T u 'R001 ATTACHPENT 27 PAGE 31 Or 41 4EAC; CRC 00tANTSfS?EM BASES' LOW TEMPERATU8E OVERPRESSURE; PROT!r ION Montinueo) and falve Cot *ing, in$trJett' uncertainties, and single failure, f;r tte tr3r. sients noted, the resulting pressure will not eaCPeo the acminal 16 Effvetive Full Pu.vr years (EF N ) Accendia G teactor vessel HDT limits ano tse ':rces ge-- Orated due tc PORV cycling do not e.stveo #0RV oiping anc Jtructural mitations. To (nsure that frati and heat input transietits 'hore severe than those af,tumed cannot occur. Technical $ptcificat hns require tre lockov of all e safety injec. ion pumps and one enaeging pump whila in H00h5 4, 5 and 5. to the reactor a vess11 head installeo, and disallow start of an RCP if secondary temperature is n re than 50'F abovt pt imary temperature. t opnratio',below 150'F out greater than 325'F with enarging and safety injec-tion pueyf 0FERABLE % allowed for up to 4 hours. Given tne snot't ti ne durat, ion that this coidition is allowed in!tietion of both trains of ufety injection curing this 4 hour time frame due tJ opirator trror er a single f s11ere occurring during testi q of a redundant channel are not considered to ce creciole accidt nts. Plant specific analysis has shown that the Cold Overpressure Mitigation System (CCMS) arming temperature may be redut ed f rom 350*F ta J20*r if the follcuing additional restricth ni are met: 1. At least one *etetor coolant pump must be i* operation. 2. Pressurip.or level is less than or equal to..'.. 3, The plant hiatup rate shall be limited to 6(;* in any one nour r,eriod. Deseconditionsappl{FbutallRCScololegsartwhentvor t'ia temperature of o iegs is less than 350 grtater tnan or eoual to 320'F. ahe6 any of the RCS RCS cold 'eg temperatures drop below 320*F, the original o rWQuirements on low temperatare operation apply, The Maxi.ma I.11c' sed PORV t,etpoint for the LTOPS will be updated cased on the i l results of examinations of reactor vessel material irradiation surveillance l specimens performed as required t,y 10 CFR Part 50, Appendix M, and in accordance with-the schedule in Table 4.4 2. l 3/4.4.9 ST Q L INTEGRITY The inservice inspection and testing programs for ASME Code Class 1, 2, and 3 components ensur.g.that the structural integrity and operational readiness of these components will be maintained t.t an acceptable level throughout the life of the plant. These programs are in accordance with Section Al of the ASME Boller and Pressure Vessel Code edition and applicable Addenda as required by 10 CFR 50.55a(g) except where specific written relief has been granted by the CommissiCn pursuant to 10 CFR 50.55a(g)(6)(1). Components of the Reactor Coolant System were designed to provice access to permit inservice inspections in accordance with Section x1 ot the ASME BoilerandPressureVesselCode,h986,JEdition. COMANCHE PEAK UNIT 1 B 3/4 4-14

l 5 Txx.92001 ATT ACHPENT 27 PAGE 32 Of 41 3/4 $ !v!QSEN:- ::DE ::CL:NG "d*!*5 1ASES 3 /4 5.. 20:U % 27:45

- e ::!uB:.: of each Reactor C:olant System (RCS) accum lat:r u

easses that a suf":ie-; <clute of onrate:.ater will oe immeaiately 'orcea +nto the eactar : ore tNough eacn of t he 601c leis in the event tne RCS oresve 'alis ceiow the cressu e cf tae acc.mulaters. This initial surge of ater nto t*e r core ord t:es the initial cooling +ec'a'n sm cun ng large RCS Dice ruotares. The limits on accumu1Ttor volume, toren concentration and pressure ensure that the assumptions usec /or accumulatcr injection in the tafety analysis are TherequiredindicatedOccumulatorvolumesandpressuresinclude [5)et m phr ent measurement uncertainty. The indicated accumulator volue s of a nd[,6 respectively, plus 4@the analjtical lin.its of [6119] gallons and[659,7] gallers, are based on . tann tolerance. The accumulator pow' operated isolation valves are considered to te " operating bypasses" in,w sentent of IEEE Std. 279 1971, wnien recuires that oypasses of a protectiv-.snction De reinoved automatically whenever :ermiss v6 conditions are nut met. In addition, as thest occumulator isolation valvas fail to meet single failure criteria, removal of power to the valves is required by BTP !CSB 18. This is accomplished via key-lock control tear: cut-off switcnes. The limits for cceration with an accumulator ino:teable for any reason except an isolation valve closec minimizes the time e,;osure of the plant to a LOCA event occurring concurrent with failure of an 3:ditional accumulator .hich may result in unacceptable peak cladding temperatures. If a closed isolation valve cannot be immediately opened, the full capability of one accumulator is not available and prompt action is required to place the reactor in a mode where this capability is not required. F4. 5. 2 and 3/4. 5. 3, _ECCS SUBSYSTEMS The OPERABILITY of two independent ECCS subsystems ensures that sufficient emergency core cooling capability will be available in the event of a t,0CA assuming the loss of one subsystem tnrough any single failure consideration. Either subsystem operating in conjunction with the accumulators is capable of supplying sufficient core cooling to limit the peak cladding temperatures within acceptable limits for all postulated break sites ranging from the double ended break of the largest RCS cold leg pipe downward, In adcition, each ECCS subsystem crovides long-term core cooling capability in the recirculation mode during the accident recovery period. I With the RCS temperature below 350*F, one OPERABLE ECCS subsystem is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the limited core cooling requirements. The limitation for a maximum of two charging pumps to be OPERABLE anc the requirement to verify one charging pump and all safety injection pumps l COMANCHE PER - UNIT 1 B 1/4 ;-1

T U.92001 ATTACHMEtiT 27 PAGE 33 0F 41 EuEqENC<::GE ::C U NG 5' STEMS BA$E$ ECS$LB$r$*!MS(Continued) to : e inocersole below 350'F peceides assurance tw a reass additie r eawe trenstant can ce relievec by the operation of a sine,le 00RV. The requirement to remove power from certain valve uDerators 11 ' sc r e:- ce.ita Branch Tecnnical Position IC$8-19 for valvss that f'il to w et ; 'e allurJ consicerations. Power is removeo via key 1c6 swt ch i on tne coi,s. i ocard. The Survel114nce Requirements provided t.o tenn.e 1PEPAB l.l.'s of eNn component ensures that at a minimum, the assunctions uPd in the saf ety analy es are met and that subsystem OPERABILITY is mainta:ned. fer throttle valve position stops and flow t Surveillknee Reovir w ts that proper ECCS flows will be maintained ir,'alante testing provide assuraMe t e event of LOCA. Mairtenance of proper flow resistance and pressure drop in the pipit g,ystem to esen injection point is _ necessary to: (1) prevent total pump fiow fra eceodin i runout conditions when the system is in its.ninimum rasistance con.i,urt',lon, (2) provide the proper flow split betwest, fi.jaction points in 6:cordarite vth-the assumptions used in the ECCS LOCA analyses, snd (3) p o ide an acce:t' ele level of ts;al ECCS flow to all injection points e ual to ce above tnet aussn.c in the ECCS LOCA analyses. 3/4.5.4 REFUEll_NG WATER STORAGE TANK The OPERABILITY of the refueling water storage ti K ( AST) is part of the ECCS ensures that a sufficient supply of beatn water is av:'ilaole for injec-tion by the ECCS in the event of a LOC 4 ine limits on. RW!T ninimum voicme anc boren concentration ensure that: (1) sufficient water ;s available witnin containment to permit recirculation cooling flow to the et 'e, (2) for small creak LOCA and steam line breaks, the rerctor wt'l remain s deritical 1 the cold condition following mining of the AST a,td the #CS vatar volumes witn all control rods inserted excipt fcir t'io most reactive c(ntrol assembly, (3) for large break-LOCAs, the reactor ull) remain sut, critical in the cold conditinn following mixing of the RWST and the RC5 rater volenes with all shutcown and i control rods fully withd Nwn, and (4) w fficiant time is available for the operator to take manuel action and complete switchover of ECCS and containment spray suction to tne containment sump without emptying the RWST or losing suction. The required indi_cated _ievel includes a[4} percent measurement ecertatsty, an unusible volume of[45,494; gallons and a required water volume F[428,43L gallbrs.* i-l The limits on Indicated water volume ard boron concentration of the RWST n!so ensure a long term pH value of between 8.5 and 10.5 for the solution recirculated wittien cnntainmnt af ter a LOCA. This pH band sAinimizes the., evoluticn ;f iocin6 and thinimizes the effect of chloride and caustic stress l corrosion ci sechanical systeils and t.omponentt. L, COMAN Fi PEA L UNIT 1 ___ . l VO M.. - -- - - - - 2.

l ATTACHMENT 27 ::s a:wEs

5*EWS PAGE 34 0F 41 BA$!i

_3 /4. 5 1. A INTERNAL PRES $URE The limitations on containment internal pressure ensure that: (1) the containmeat structure is prevented from exceeding its design negative pressure differential of 5 esid with respect to the outside atmosonere, and (2) tne 0ntaitreat peak pressure does not exceed the design pressure of 50 psig caring a LOCA. Theindicatedcontainmentpressurevaluesof60.3psigand1.3psig correspond to analytical limits of[-0.5]psig and[1.5)p],ig, respe{tti]vel allowance for measurement uncertainty. 48 3]psig which is less than design pressure and is egnsistent with (.safety analyses. This value includes the limit of[1.5jpsig for initial positive containment pressure. 3/a 6 1.5 AIR TEMPERATURE The limitations on containment average air temperature ensure that the overall containment average air temperature does not exceed the initial tem-Jeanure condition assumed in the safety analysis for a LOCA or steam line creak accident. at least 2 of the measurements made at the listed locations, by fixed o portable instruments with allowance for temperature measurement uncertainty. 3/4 6.1.6 CONTAINMENT STRUCTURAL INTEGRITY This limitation ensures that the structural inte. ity of the containment will be maintained comparable to the original design standards for the life of the fac'14t. Structural integrity is requ red to ent.re that the containment will withstand the maximum pressure of[48. psig in the event of a LOCA. A visual inspection in conjunction with tte ype A leakage tests is sufficient to cemonsteate this capaDility. 3'a 6.1.7 CONTAINMENT VENT!LATION SYSTEM The 48 inch and 12 inch containment and hydrogen purge supply and ennaust isolation valves are required to be locked closed during plant operations since these valves have not been demonstrated capable of closing during a LOCA or steam line creak accident. plant operation ensures that excessive Quantities of radioactive materialsMaint will not be released via the Containment Ventilation System. To provide assurance that these containment valves cannot De inadvertently opened, the valves are locked closed in accordance with Standard Review Plan 6.2.4 which includes mechanical devices to seal or lock the valve closed, or prevents power from baing supplied to the valve operator. The use of the Containment Ventilation System during operations is restricted to the 16-inch pressure relief discharge isolation valves (with an effective diameter of 3 inches) since, these venting valves are capable of closing during a LOCA or steam line break accident. Therefore, the Exclusion Area dose guideline of 10 CFR 100 would not be exceeded in the event of an accioent during containment venting operation. COMANCHE EaK - UNIT 1 8 3/4 6-2

TXX 92001 ATTACHMENT 27 PAGE 35 0F 41

NTA!NutV i<5'EWS BASES CON?AINMEh? VENT!LA'MN SYSTEM (Continued)

Leaka*e iMegrity tests with a maximum allowable leakage rate *or cortain. ment venti 14tien valves will provide early indication of resilient material seal degracation and.ill allow opportunity for repair before gross leakage fatiures could d&>eloo. The 0.60 L, leakage limit of Specification ).6.1.20. sneil not be exceeced = hen tot leakage rates determined by the leakage integrity tests of these valves are added to the previously determined total for all valves and eenetrations subject to Type B and C tests. 3/4 6.2 DEPRESSURIZAT!0N AND C00Ll40 SYSTEMS 3/4 6.2<1 CONTAINMENT SPRAY SYSTEM The OPERABILITY of the Containment Spray System ensures that containment deoressuritation and cooling capability will be available in the event of a LOCA or steam line break. The pressure reduction and resultant lower contain-ment leakage rate are consistent with the assumptions used in the safety anclyses. The Containmert Soray System which is composed of redundant trains, pro-d ces cost

accident cooling of the containment atmosp*ere.

However, the Con-tainment Spray System also provioes a mechanism for

v. oving iodine frem the containment atmosp9ere and therefore the time require tats for restoring an inoperable Soray System to OPERABLE status have been tintained consistent with that assigned etser inoperable ESF equipment.

3<4.6.2.2 SPRAY a:0 ':vE SYSTEM The OSERABILI't cf the Spray Additive System ensu es that sufficient NaOH is added to the containment spray in the event of a LOCA. The limits on NaOH volume anc concentration ensure a lo.$g term pH value of between 8.5 and 1Q.5 for the solution recirculated within containment after a LOCA. This on band minimizes the evolution of iodine and minimizes the effect of cnlorice and caustic stress corrosion on mechanical systems and components. The contained solution volume limit includes an allowance for solution not usaole because of tank discharge line location or other physical characteristics. These assump-tions are consistent with the iodine removal efficiency assumed in tre safety analyses. The reovirec 'ecicated level band of[9h tol94)lgallons to[5314 for the Spray Aod tive i Tank corresponds to a9 analy jcal limit band of[490Q

gallons, respectively, ano iccisces a 3.3f[. measurement uncertainty.

CO*ANCHE DEAK - UN

1 B 3/4 6-3

VXX-92GM ATTACHMEhf 27 PAGE % 0F 41ptaNT sy$'EuS EASES 3'471.2 aux!LIARY FEE 0 WATER SYSTEM The OPERABILITY of the Auxiliary Feedwater System ensures that the Reactor Ccolant System can be cooled down to less than 350'F from normal coerating conditions in the event of a total less of offsite power. Each electric motor driven suiiliary feedwater pump is capable of celiver-ing a total feedwater flow of 430 gpm to two steam generators at a pressure of 1221 psig to the entrance of the steam generators. The steam driven auxiliary feedwater pump 15 capable of delivering a total feedater flow of 860 gem to four steam generators at a pressure of 1221 psig to the entrance of the steam This capacity is sufficient to ensure that adequate feedwater flow generators. is available to remove decay heat and reduce the Reactor Coolant System temp-erature to less than 350'F when the Residual Heat Removal System may be placed into operation. The Auxiliary Feedwater System is capable of delivering a total feedwater flow of 430 gpm at a pressure of 1221 psig to the entrance of at least two stetm generators while allowing for: (1) any possible spillage through the cesign worst case break of the main.feedwater line; (2) the design worst case single failure; and (3) recirculation flow. This capacity is sufficient to ensure that adequate feedwater flow is available to remove decay heat and reduce Reactor Coolant System temperature to less than 350*F at whien point 'e Resioval Heat Removal System may be placed in operation, The test flow ror the steam-driven auxiliary feedwater pump at a prsisure of greater than or equal to 1450 psid ensures this capability. The auxiliary feedwater flow path is a passive.f';w Dath based on the fact that valve actuation is not requirec in order to supp' tiow to the steam generators. The automatic valves tested in the flow path are the Feedwater Split Flow Bypass which are required to be shut upon initiation of the Auxiliary Feeowater System to meet the requirements of the accident analysis. Both steam supplies for the turbine-driven auxiliary feedwater pump must te OPERABLE in order to meet the design bases for the complete range of accident analyses. The allowed outage time for one inoperable steam source is ;onsistent with tre lower probability of the worst case steam or feedwater line creak accident. 3/4.7.1.3 CONDENSATE STORAGE TANK The OPERABillTY of the condensate storage tank with the minimum water volume ensures that sufficient wate-is available to maintain the RCS at HOT STANDBY conditions for 18 hours with steam discharge to the atmosphere concurrent with total loss-of offsite power or 4 hours at HOT STANDBY followed by a cooldewn to 350'F at a rate of 50'F/hr for 5 hours. The contained water volume limit incluces an allowance for water not usable because of tank discharge line loca-tion or other physical characteristics. The required indicated level includes a 3.5 percent measurement uncertainty, an unusable volume of[12,000] gallons ano a required usable volume of[250,000] gallons. NUREG-0737. Item II E.1.1 requires a backup source to the CST which is the CPSES Station Service water System, which can be manually. aligned, if reauirea in lieu of CST minimum water volume. COMANCHE PEAK - UNIT 1 B 3/4 7-2

TXX-92001 ATTACHMENT 27 PAGE 37 0F 41 a'aN' 5 5 tu! iAJU u! A *ta:(2A'URE MON!?C8tN3 (Contirued) f!PrER A*(.Rt.:4!f ('O Normal Acncrmal ar,e Concie ces

enait ons Area mom te ee
t0M D'et*: rm 40 143 General Area Barrier CROM Shroco Enhaust teacter Casit 135 175 Reactor Cavity innaust Detecter wel'y R.C, Pioe Penetration 200 209 General Areas innacst (N*1i Reacter Cavitf Eanaust Cetecters) 2/4.7.11 JPS dVAC $Y$*(4

-{ ike CP[RABll,;!Y of the UPS HVAC System ensures that the uninteruptible cc er sucoly ano :istricution reces ambient tir tericeratures do not exceto the aliowaole terceratures cea $cecification 3/4.7.10 for ntinuous-atty st' g s 'or tne easicment and instrumentation Cooled by th s .sioment. CCMANCHi ik UNIT 1 8 3/t. 7 7 ) _m___-___-_ ^ ~ ^ ~

Txx.92001 ATT 4CH"ENT N j PAGE 38 Of 41 g3 :t:,t.:Ns C:t:c::ss 345Ei )491 9:0C4 CCNCENtcp*;N The I'mitatdoas on reactivity conoitions during RE%ELING ensure tant; ( *. ) t*e reactor.tii remain suteritical curing CORE ALTERATICNS, arc (2) e uni'erm toren concentration is mainteired for reactivity control

e.ater

.cipe nhing cirect access to the react:r vessel. These limitati;es see corlistent with the initial conditions assutec fer the coron ciluti:n ine' cent in the safety analyses. Tne value of 0.55 or less for (,gf inclu es a 3 ak/k conservative allowance for uncertainties. Similarly, the toren concentration value of 2000 ppm or greater includes a conservative urcertainty 'iowanceof[$hppmboron. The locking closed of the required valves ouring neling operations precludes the possibility of uncontrolled boren cilution of the filled portion of the RCS. This action prevents flow to the RCS of untorated water by closing flow ;aths f rotn sources of unboratec.ater= 3/4.9.2 lNSTRUMENTATICN The OPERABIL!iY of the source range reutron flux monitors ensores t'at reduncant monitoring capability is availaDie to detect changes in the reactivity concition of tne core. 3'4.9.3 DECAY TIME The minimum requirement for resctor subcriticali-prior to movement of ireadiated fuel assemblies in the reactor vessel ensures tnat sufficient time hat elapsed to allow the radioactive decay of the short lived fission products. This decay time is consistent with the assumotions used in tre 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 will be restricted from leakage to the environment. The OPERAB!LITY and closure restrictions are sufficient to restrict radioactive material release from a fuel element rapture based upon the lack of containment pressurization potential while in the REFUELING MODE. 3/4.9.5,CCMMUNICATIONS The requirement for communications capability ensures that refueling station personnel can be promptly informec of significant :hanges in the facility status or core reactivity conditions during CORE ACTERATIONS. OCMA% CME DEAN. UNIT 1. RhA*

i TXX-92001 ATTACHMENT 27 PAGE 39 0F 41 50 DESIGN FEATURES 3 5.1 $!TE M'USIONAREA 5.1.1 The Exclusion Area shall be as shown in Figure 5.1 1. LOW DOPULA':0N ZCNE 5.1. 2 The Low Population Zone sns11 be as shown in Figure 5.1 2. MAP CE/lN!NG UNRESTRICTED AREAS AND $1TE B0UNDARY FOR RADI0 ACTIVE GASE0 LIOu!D EF hvENT5 5.1. 3 Infort.*ation regarding radioactive gaseous and liquid effluents, which will allow identification of structures and release points as well as defini-tion of UNRESTRICTED AREA 5 within ths $1TE BOUNDARY that are accessible to MEMBERS OF THE PUBLIC, shall be as shown in Figure 5.1 3. fhe definition of UNRESTRICTED AREA used in implementing these Technical $pecifica:;ons has been expanded over that in 10 CFR 20.3(a)(17). The UNRESTRICTED AREA boundary may coincide with the Exclusion Area Boundary, as cefined in 10 CFR 100.3(a), but the UNRESTRICTED AREA does not include areas i over water bocies. The concept of UNRESTRICTED AREAS, established at or beyond the SITE BOUNDARY. is utilized in the Limiting Conditions for Operation to keep levels of radioactive materials in liquid and gaseous effluents as low as is reasonably achievable, pursuant to 10 CFR 50.36a

5. 2 CONTAINMENT CONFIGURATION 5.2.1 The containmeat >uilding is a steel lined, reinforced concrete building of cylindrical shace. with a dome roof and having the following design features:

Nominalir,sidediameter=[135

feet, 3.

b, ominal '. side heignt = 192.5, feet. (Dome 67.5 feet; total = 260) feet) Nominal thickness of concrete walls ={4.5[

feet, c.

Nominal thitkness of concrete roof =(2.5} feet, d. Nominalthicknessofbasemat=12.0] feet. e. f. Nominal thickness of steel liner wall = 3/8 inen. (Dome = 1/2 inch. Base Mat r 1 s inch), and g. Net free scluee =, 2.985.000 cubic feet. 4 E 1 C;MANCH[ D[AK a UNIT 1 51

Tx X - 92001 ATTACHtiENT 27 i' ACE 40 OT 41 'E5'.;N ia*,*Ei <CLlur 5.4.2 ' e total. ate-an: steam volute of t*e Reac.or ;oolant 5 stem

2.'35 * '".'

0.0'* 'te*, at I comi"ai I of h39. 'I. i avg

a. t.

. r. - :. =. ;.... ;. -, :. ;.. a r. ; - *.. y 5.5.'

e :

s

ete:re':q i to.er

~ ';.re 5.1 L s ai' :e io:stes ss sm:.n

5. 6 t'.'E.

5 7C D AG E CRITICALI'Y 5.6.1.1 The spent fuel storage racks are designec anc shall be maintaiae: .it - a

ff acu'. nle-t t: 'ess tran or ecual o 0.i$ when floo:e:

a. 1.* uncoratec =ater, which 1.ncluu.s a conservative allowance 'or vn:ertaint es as cescrioec in tion 4.3 of the F5AR, anc i t. A cominal 16 inch center-to center cistance tet een fuel assetoises placed in :ne storage racss. 5.6.'. '5e s,,, ' : n e a fuel 'or the first core loa: g stored cry in :ne scant fuei storage acxs sna'1 not exceed 0.98 wnen a. ecus foam tocerati:n 4s assume. 04AINAGE 5.6.2 ine spent fue! storage cool is cesigned ano shall be maintai e: to

revent inaevertent craining of the pooi telow elevation 854 feet.

CAdddTt*>= 5.6.3 The two spent fuel storage pools are cesigned and shall et caintainec with a storage capacity limitea to no more than 1116 fuel assemolies.

5. 7 COMPONENT C)CLIC OR TRANSIENT LIMIT 5.7.1 The comoonents identitfee in Table 5.7-1 are designec and shall te maintainec witnin tne cyclic or transient limits of Taole 5.7-1.

1 COMANCME DEAK - UNIT 1 _ _ _ _ - - - _ _ _ _ _ - *6

VXX-92001 Af?ACHMENT 27 PAGE 41 0F 41 ADMINISTRAT!yECONTROLS CORE OPERATING LIMITS REP 0RT (Continued) WCAP 10216-P A, " RELAXATION OF CONSTANT AXIAL OFFSET CONTROL FQ SURVEIL- \\ LANCE TECHNICAL SPECIFICATION," June 1983 (W Proprietary). (Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor (W (z) surveillance requirements for F Methodology).) q WCAP-8200, "WFLASN, A FORTRAN-IV COMPUTER PROGRAM FOR $1MULATION OF TRAN-SIENTS IN A MULTI-LOOP PWR " Revision 2, June 1974 (i Channel Factor W Proprietary). (Methodology for Specification 3.2.2. - Heat Flux Ho ' WCAP 9220-P-A,(W Proprietary)." Westinghouse ECCS Evaluation Model, F February 1978 (Methodology for Specification 3.2.2. - LHeat Flux Hot CKannel Factor.) I t The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met. The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Oesk with copies to the Regional Administrator and Resident Inspector. SPECIAL REPORYS 6.9.2 In addition to the applicable reporting requirements of Title 10, code of Federal Regulations, special reports shall be submitted to the Regional Administrator of the Regional Office of the NRC within the time period speci-fied for each report. l 6.10 RECORO RETENTION 6.10.1 In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated. j A.10.2-The following records shall be retained for at least 5 years: ) a. Records and logs of unit operation covering time interval at each power level; b. Records and logs of principal maintenance activities, inspections, repair, and replacement of principal items of equipment related to nuclear safety; 1 COMANCHE PEAK - UNIT 1 6-20a Amendment No. I, 6 ,~ ....__u.4,- m ._m, ..4 m ,g

Txx 92001 ATTACHMENT 28 PAG ( l of 1 Several lechnical Specificatica changes are cur rent ly being considered f er CISLS Units 1 and 2. Many of these changes are not directly attritutable to the licensing of Unit 2 and are out included in Attachments I through 26. It is however, desirable that CPSE5 linit 2 be licensed incorporating many of t hew changes. License Amendment Requests to Doclet No. 5,0 445 (CPSEL Unit 1) will be sut,mitted in the near future on the subject areas listed below if approved by the TV Electric internal review and approval rrocess: A. Generic Letter 91 08. Pemoval of Components Lists from Technical Specificatiens, for containment isolation valves and containment penetration conductor overcurrent protective devices. B. Removal of the Reactor Vessel Material Surveillance Program Withdrawal Schedule in accordance with the gui1bnce of Generic Letter 91 01. C. A revision t o the RCS total flow r ate surveillance (4.2.5.4) to specif y a minimum power fer performance of this measurement. D. Deletion of allowance on the qualifications of the Assistant Radiation Protection Manager in Section 6.3,1. E. Addition of a 3.0.4 enception which currently exist in the standard Technical 5pecifications for specification 3/4.J.4. f. A change to the Cont rol Room HVAC System (3/4.7.7) to credit operation with 3 independent 50% capacity control room HVAC systems. G. Deletiori of testing required 'during shutdcwn" or during the ' REFUEL!flG MODE' that is already being performed during " POWER OPERATION' H. A revision to ACTION 28 (3/4.3.3.1) to tale credit for the additional Control Room air intake radiation monitor added f'or each air intale. 1. Clarification to the ACTIONS to be taken upon failure to achieve a positive pressure in the control room, J. Revision to Table 4.3-1 Notation 6. to allow perf ormance of the Incore-Excore calibration at power levels below 75% of RATCD THERMAL POWER. K. Generic Letter 91 13, (Generic Issue 130), essential service water system f ailures at ino ti-unit sites. l t l}}

ML20086U414

Text

~. ~ :

.

Navigation menu

Search