Text

Proposed Ts,Relocating cycle-specific Parameter Limits to COLR
	ML20127G275
Person / Time
Site:	Mcguire, Catawba, McGuire
Issue date:	01/13/1993
From:	; DUKE POWER CO.
To:	;
Shared Package
ML20127G266	List: ML20127G264; ML20127G275;
References
	NUDOCS 9301210204
	Download: ML20127G275 (135)
	v • d • e

..

C-L 3

Changes to Technical Specifications (Catawba) i

.I

)

-i i

[jjl21020493h((j69 2-1 I

p ADOCK 05 PDR

-un f++-

l n4 TABLE 2.2-1 (Continued) g TABLE NOTATIONS NOTE 1:

OVERTEMPERATURE AT 3(P - P') - f (AI)}

f1

[I (1 + r s) -

+

3 AT (1 + r35) I O

- 2 7 3 e

1 o

w Heasured AT by Loop Harrow Range RIDS; Where:

AT

=

g.

1 + r,5 lead-lag compensator on measured AT; N

=

y, g

Time constants utilized in lead-lag compensator for AT, :, = !? r,

=

12

~ ? E;-

L a%,,ls Rey,et; 13, as presenied on the Cere O eenfa%3 f

a 1

Lag c mpensator on measure 4 4T:

=

5 1+13 Time constant utilized in the lag compensator for AT, : - 0; as presented in l

j 3

=

13

+he. Core OpersfQ L;,nlts Repart; g

4 s

Indicated AT at RATED THERHAL POWER; AT

=

f Repd; a

-1.1"53 - Overtemperd-ce AT reader fdp seip:s1 as prese,&ed n ik Care O ermia G,,

ey p

y

~

K,

=

ke<fup seI o:nt pe >ahy toefflcle.t as K

= 931G3l*i Ovaleeperdure Ki~re eforfr:p p

P'c'* ~+d l~

C*<

O *'d;'s Li~fr R*Pi 2

0 1+rs= The function generated b/ the lead-lag compensator for T,yg pg y,

3 dynamic compensation; gg I, - ?? s, Time constants utilized in the lead-lag compensator for T,

=

13 14, pre Scaled na the Cort Operd!*3 li,a,'h Rep W ;

gg b - ' C; as RE Average temperature, *F; T

=

.o $

Lag compensator on measured T,

=

"."[

y, 3

E5 3;

Time constant utilized in the measured T, lag compensator, q - 0; as presenhJ

=

is Llm.'t> Report; Coce. Opedk3 mg in itse

~

4

4 c,

UN!! 1 l

f 4

' c*o TABLE 2.2-1 (Continued)

TABLE NOTATIONS (Continued) a j

E NOTE 1:

(Continued) i Z

I'

- 590.8 F (Nominal T allowed by Safety Analysis);

avg

'K

= Do141+; overte,.,perature AT read.e te;p depressnes% f:en se+polnt peanally coeff;cle f as l e

presea,ted in the Cace Opeeat: y Li,.h Repor t;

.i 3

P'ressurizer pressure, psig-I P

=

2235 psig (Nominal RCS operating pressure);

P'

=

i r

5-

= Laplace transform operator, s 8; and f (AI) is a function of'the indicated difference'between top and bottom detectors of the

~

power-range neutron ion chambers; with gains to be selected based on measured instrument p.

response during plant STARTUP tests such that:

L;,*t, RIet;i betweed oslfke<ad *nec.ative.* f(AT) bre&: ts as prese. fed ir, }/,e Core O m

c tAe 0>

30.0% =d d.0%,

l c

(i) for q

-q t

b

.y f (AI) = 0, where q and q are percent RATED THERMAL POWER in the top and bottom 3

b is total THERMAL POWER in percent of halves of the core respectively, and qt'Nb RATED THERMAL POWER;

'the {(br) *negal:ve " bred olef prese.ded h k Core. O *r'Yl~3 LAlhdeptq f

f is a re negative than -30.0%, the

. j!yy (ii)

For each percent AI that the magnitude of q -

b

-q AT T rip Setnoint shall be' automatically reduced by 910 cf aTo; am t.

yg 14e f(4T) *y lve" S ope prose sted in k Ge.()pdQ 4;a.lts Qt ;

f l

and j

b is m re Positive than e3.0%, the aT Trip -

gg

-(lii)

For each percent AI that the magnitude of q g

Setpoint shall be automatically reduced by 4 416E cf AT.

o i

M i

i

'The-channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by 11 NOTE 2:

l more than 3.0%.

UC 14e $ (at)

fos: Nae ~.S/ ope presented in Ac ((hl)"posll0e? hterk a:Zpreseated f

in Hg Core Oper.]lng L;,,,lts Report 9'

Ae Core. ofeat.b Ll,,,lh Rep,,f

un:T 1 l

3 TABLE 2.2-1 (Continued)

TABLE NOTATIONS (Continued)

E NOTE 3: OVERPOWER AT U

of (1 + T,5) (

1 II S

)

I I

}

I

- T"] - f (al))

(1 + r25) (1 + T331.~

0 II * '7S)

(l + s) T - Ks [T (1 + tsS)

T i lK

-K 2

4 e-N As defined in Note 1, Where:

AT

=

I**0 As defined in Note 1, 1 + r25 As defined in Note 1

=

13, t; fg As defined in Note 1

=

3 so As defined in Note 1

=

13 y

L As defined in Note 1, AT,

=

Overfoar~ LT reuk tr'p.setp'd as presented inKe Cere Open C~h hf 1.0819 K

=

4 0.02/*f for increasing average temperature and 0 for decreasing average gy K

=

3 gg temperature, EE t'S The function generated by the rate-lag controller for T,,g dynamic

=

gg 3

compensation, 55 Time constant utilized in the rate-lag controller for T,,g,

, - 10

, as gesede)

=

17 In h Orc M.'=, C-.h Reg

$S 1

As defined in N te 1,"

=

l + rGS

^^

CC 33??

As defined in Note 1,

=

is my vu 4

o

-UNH f 6

TABLE 2.2-1 (Continued)

TABLE NOTATIONS (Continued)

(

"F N

T P

m se care opa% t,.-.s s,".e

' F ##

E NOTE 3:

(Continued) 001291#F-f or T > 590.8*F and K. = 0 f or T < 590.8*F,

l i

K.

=

As defined in Note 1, T

=

w Indicated T at RATED THERMAL POWER (Calibration temperature for ni 1"

=

instrumentation, < 590.8*F),

l As defined in Note 1, 5

=

and f 3 a function of the indicated differences between top and bottom detectors ofth$p(al)ower-range neutron ion chambers; with gains to be selected based on measured instrument response durir.g plant startup tests such that:

c t e ~ pes %ls.4 *eegEke? fn(hn buskjndois as))resesled o~.% Core Oyemi:y m

h q between* 35% a..d 235% "I; f2(I = 0, where q and g are percent O

(i) for q 3

G RATED THERMAL POWER in the top and bottom halves of the core respectively, and q

+q is total THERMAL POWER in percent of RATED THERMAL P0hTR, mf;.2 L.%:fr /?ept the IloE)" ne

nbre ci

rese ded *'n 14 Care re negative than )

n (ii) for each perce AI that t e ma nitude of q

-g is b

-35% AI, the AT 1 rip Setpoint shall be automatically reduced by 7. 5 cf AT.; and

?$

4te fs(AA *neyfbe. Slo preseded in ike Cere,0persf Dalh hacet }

,3 gg (iii) for each pe cent al that magnitude of q N

E#5 t

b sa

g

+354-al, the AT Trip Setpoint shall be automatically reduced by 7.5 cf a!..

The $(AQ " pas Net' brea kpc!# presedeA E., he (m. O eruh Liarf3 hpo,#

f 3

NOTE 4:

The channel's maximum Trip Setpoint shall not exceed its computed Trip setpoint by more than 2.8%.

y9 Ye $bD 'f=U*bcWafe.,oresented k ike [ere. }rd,%3 4.'r h Syf

~

n I

D

C -0 1

l REACTIVITY CONTROL SYSTEMS 1

1 BORATED WATER SOURCE SHUTDOWN LIMITING CONDITION FOR OPERATION 1

3.1.2.5 As a minimum, one of the following borated water sources shall be OPERABLE:

l a.

A Boric Acid Storage System with:

1)

A minimum contained borated water volume of-lh000-gaHens, l

)

2)

A minimum boron concentration of$7000 pp petJ 1

a s pre se n'. e d in th e Cere. a eeJ;n Lim.h Re r

and as pre.sente d in the Cae OpmtQ Ll~,h R.,bi'>

i 3)

A minimum solution temperature of 65'F.

l b.

The refueling water storage tank with:

1)

A minimum contained borated water volume of-4n000-ge44ent, l

us presenteGk Core. Cher.1% U % R ut J 2)

A minimum boron concehtration ef-G and as preeniedin N Core o nih L;m.h Ryert n

3 3)

A minimum solution tbmperature of 70'F.

APPLICABILITY:

MODES 5 and 6.

ACTION:

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

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

a.

At least once per 7 days by:

1)

Verifying the boron concentration of the water, 2)

Verifying the contained borated water volume, and 3)

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

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the refueling water storage tank temperature when it is the source of borated water and the outside air temperature is less than 70*F.

CATAWBA - UNITS 1 & 2 3/4 1-114 Amendment No. 82 (Unit 1)

Amendment No. 76 (Unit 2)

I c '1 1

l f

REACT 1VITY CONTROL SYSTEMS i

j BORATED WATER SOURCES - OPERATING l

LIMITING CONDITION FOR OPERATION

.~

)

3.1.2.6 As a minimum, the following borated water source (s) shall be OPERABLE i

as required by Specification 3.1.2.2:

i l

a.

A Boric Acid Storage System with:

1)

A minimum contained borated water volume +f--?2,000 ;:11:n:,

l

}

J aa presented I., M Cse op.ca% 4.' as Ry:,hnd i

2)

Ame' a be.)ron concent'ratioW of 7000 J a

f my

.ma k +be Gee Opraky 4,%lh Rot *#

3)

A rie.

wa solution temperature of 65'F.

i j

b.

The r,efueling water storage tank with:

mammm j

1)

A contained borated water volume of-:t ic::t 253,512 ;;1 h n;, t 3

2)

A minimum boron concentration of 2000 p bb as pre.sestedia fie bre. 0perd.y L'nh R. c1 },

e' '%

3)

A minimum solution temperature of

'F, and

{k

.n i

4)

A maximum solution temperature of 100*F.

'4 C'A

j APPLICABILITY:

MODES 1, 2, 3, and 4.

I ACTION:

9 0 a.

With the Boric Acid Storage System inoperable and being used as one 3"

of the above required borated water sources, restore the system 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 within h

the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and borated to a SHUTOOWN MARGIN equivalent to at y

least 1% ak/k at 200 F; restore the Boric Acid Storage System to

1 OPERABLE status within the next 7 days or be in COLD SHUTOOWN within d

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

b.

With the refueling water storage tank inoperable, restore the tank W

i to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY

}A within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTOOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

3 S

%1 O

E

$t 1'

f i

27 4

- CATAWBA - UNITS 1 & 2 3/4 1-17 Amendment No. 82 (Unit 1)

- Amenoment No. 76 (Unit 21

C-P INSTRUMENTATION BORON DILUTION MITIGATION SYSTEM l

s LIMITING CONDITION FOR OPERATION

.l 3.3.3.11 As a minimum, two trains of the Boron Dilution Mitigation System j

shall be OPERABLE and operating with Shutdown Margin Alarm ratios set at less than or equal to 4 times the steady-state count rate.

APPLICABILITY:

MODES 3, 4, AND 5 ACTION:

(a) With one train of the Boron Dilution Mitigation System inoperable or not operating, restore the inoperable train to OPERABLE status within j

48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , or l

(1) suspend all operations involving positive reactivity changes and verify that valve NV-230 is closed and secured within the next hour, or (2) verify two Source Range Neutron Flux Monitors are OPERABLE with Alarm Setpoints less than or equal to one-half decade (square root of 10) above the steady-state count rate and verify that the combined flowrate from both Reactor Makeup Water Pumps is less than or equal to-150-gpm-5 within the next hour. A he we{ Mode-3-or4)-or-75-gpm-(Mode b) rutw (u Mca p W e ecerrop vum n

w um pys e p en m q y,nuS (b) With both trains of the Boron Offu'ti'on Mitigation System inoperable 4

or not operating, restore the inoperable trains to OPERABLE status within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , or (1) suspend all operations involving positive reactivity changes and verify that valve NV-230 is closed and secured within the next hour, or (2) verify two Source Range Neutron Flux Monitors are OPERABLE with AlarmSetpointslessthanorequaltoone-halfdecade(square root of 10) above the steady-state count rate and verify that the combined flow rate from both Reactor Makeup Water Pumps is less than or equal to -150-gpm-(Mode-3-or-4)-or-75-gpm-(Mode-5)-

within the next hour, w e wexso r m e eu p o d er Pu mp hio m ente.

p rew nw i n w core o p erca;ne

.Li m H c Reporf-h SURVEILLANCE REQUIREMENTS 4.3.3.11.1 Each train of the' Boron Dilution Mitigation System shall be demon-l strated OPERABLE by performance of:

(a) A CHANNEL CHECK at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , CATAWBA - UNITS 1 & 2 3/4 3-85 Amendment No.10 Unit 1 l

Amendment No. 9 Unit 2

C'9 l

INSTRUMENTATION l

SURVEILLANCE REQUIREMENTS (Continued)

(b) An ANALOG CHANNEL OPERATIONAL TEST at least once per 31 days, and j

(c) At least once per 18 months the BDMS shall be demonstrated OPERABLE by:

(1) Verifying that each automatic valve actuated by the BDMS moves to its correct position upon receipt of a trip signal, and (2) Verifying each reactor makeup water pump stops, as designed, j

upon receipt of a trip signal.

4.3.3.11.2 If using the Source Range Neutron Flux Monitors to meet the require-I l

]

ments of Technical Specification 3.3.3.11, l

(a) The monthly surveillance requirements of Table 4.3-1 for the Source 1

Range Neutron Flux Monitors shall include verification that the i'

Alarm Setpoint is less than or equal to one-half decade (square root of 10) above the steady-state count rate.

(b) The combined flow rate from both Reactor Makeup Water Pumps shall be j

verified as less than or equal to 150 gpm-(Hode-3-or4)-OMS-gpm-(Mode 5)atleastonceper31 days.]

MU h") F " k' 7 u b@.c t M oup p r 9 n\\ fd M fiQ.

png h tk pers I

l CATAWBA - UNITS 1 & 2 3/4 3-86 Amendment No.103 (UNIT 1)

Amendment No. 97 (UNIT 2)

.e

--e

,,a--

C-lO 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS COLD LEG INJECTION LIMITING CONDITION FOR OPERATION 3.5.1 Each cold leg injection accumulator shall be OPERABLE with:

a.

The discharge isolation valve open, b.

A contained borated water volume of between 7704 and 8004 gallons, on concentration +f between 4900-ar.d 2100-ppe, A bog /imih preseded k JAe Core 0 erd c.

!!5and67d,rt thW Lim;h Re Anitrogencover-pressureofbet/een psig, and d.

A water level and pressure channel OPERABLE.

e.

APPLICABILITY: MODES 1, 2, and 3*,

2 h

ACTION:

tLe lower lW,t presed.,d la 14c. Core. Oper4!.3 ;m.h 2cprT 4

With one cold leg injection accumulator inoperable, except is a resul a.

of a closed isolation valve or boron concentration less than 4900 pp"r restore the inoperable accumulator to OPERABLE status within 1 hour1.157407e-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 HOT SHUT 00WN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.

With one cold leg injection accumulator inoperable due to the isolation valve being closed, either immediately open the isolation valve or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUT 00WN within the gollowina 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

g

%e loose Nmit pres.. ted in ibe Core Opers% 44 lts Report v

3

c. CWith one accumulator inoperable due to bofon-concentration less than 1900 ppe.and:

tco egyalto the lowec /Mpfad in m Q Cpd;y M Q,4 Y

1)

The volume weighted average boron concentration or the accumula-tors 4900-ppe.or greater, restore the inoperable accumulator to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of the low boron determination 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 reduce Reactor Coolant System pressure to less than 1000 psig within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

nain;emm rey;ted iv encare. Post LotA suLeriflcality ?tou<

f l

ths R

y The volume weighted average boron concentration of the de /osec /,*ailfyme.M2) in VAc. Com opb) % _ acc_umulators less than $900 pp& but greater than-1000 ppa,"

restore the inoperable accumulator to OPERABLE status or L;mlfs Repet

((0 return the volume weighted average boron concentration of the Me loser"/,%,1 presc4d accumulators to greater than 4900- ppa and in % C. ore. Opsesi: 3 L:~h Repeef

Reactor Coolant System pressure above 1000 psig.

CATAWBA - UNITS 1 & 2 3/4 5-1.m Amendment No. 101(Unit 1) 2 Amendment No.

95(Unit 2)

k C-It i

k i

EMERGENCY CORE COOLING SYSTEMS I

P ITING CONDITION FOR OPERATION (Continued)

M?

' " ' ' 'h (Continued) / 7",sa.a'"' '"c".","Y" ^ *i.."..s8.,.s ACTION:

o,ain rce enter ACTION c.1 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months of the low boron determination or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce Reactor l

i Coolant System pressure to less than 1000 psig within the fol-l lowing 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

3)

Th volume weighted average boron concentration of the accumula-tors -1000 ;; or less, return the volume weighted average boron j

concentration of the accumulators to greater than 1000 m nd enter ACTION c.2 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of the low boron determination or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce Reactor Coolant System pressure to less than 1000 psig within the follow-

{

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

-ffe m intm s m rey l red h ensure pd -Loch s,6<,,*hgl.17 A*

SURVEILLANCE REQUIREMENTS l

t i

4.5.1 Each cold leg injection accumulator shall be demonstrated OPERABLE:

)

l a.

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

1)

Verifying, by the absence of alarms, the contained borated water volume and nitrogen cover pressure in the tanks, and 4-2)

Verifying that each cold leg injection accumulator isolation valve is open.

b.

At least once per 31 days and within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months af ter each. solution l

volume _ increase of greater than or equal to 75 gallons by verifying the boron concentration of the accumulator solution;

.t l

c.-

At least once per 31 days when the Reactor Coolant System pressure j

is above 2000 psig by verifying that power is removed from the isolation valve operators-on Valves NI54A, N!658, NI76A, and N!88B i

and that the respective circuit breakers are padlocked; and t

l d.

At least once per 18 months by verifying that each cold leg injection accumulator _ isolation valve opens automatically under each of the following conditions:**

i t

1)

When an actual or a simulated Reactor Coolant System pressure i

signal exceeds _the-P-11= (Pressurizer Pressure Block of Cafety Injection) Setpoint, and 2)

Upon receipt of a Safety-Injection test signal.

- This surveillance need not be performed until prior to entering HOT STAN0BY-following the Unit-1 refueling.

=

endment No.101 (Unit 1)-

Am CATAWBA - UNITS 1 & 2 3/4 5-22-H Amenoment No. - 95 (Unit 2)-

4

C - 12.

i EMERGENCY CORE COOLING SYSTEMS i

3/4.5.4 REFUELING WATER STORAGE TANK i

i j

LIMITING CON 0! TION FOR OPERATION J

{

3.5.4 The refueling water storage tank shall be OPERABLE with:

i a.

A minimum contained borated water volume of 363,513 gallons, i

i b.

A boron concentration of between-2000 erid 2100 pr af bar:r.,

f N /;mlb preseM in n Cara. Operaw Liuh Repet J s

j c.

A minimum solution temperature of 70*F, and f

d.

A maximum solution temperature of 100'F.

APPLICABILITY:

MODES 1, 2, 3. and 4.

1 i

ACTION:

i i

With the refueling water storage tank inoperable, restore-the tank to OPERABLE j

status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be,in at least HOT STANDBY within 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 .

i.

I

+

1 SURVEILLANCE REOUIREMENTS l

4.5.4 The refueling water storage tank shall be' demonstrated OPERABLE:

a.

At least once per 7 days by:

I j

1)

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

-Verifying the boron concentration of the water.

i b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the refueling water storage j

tank temperature wnen the outside air temperature is less than 70'F or greater than'100'F.

I J

1 4

4 i

2 12 i

CATAWBA - UNITS-1 & 2 3/4 5-11

.. 1

+.

~.

l C-I3 l

REFUELING OPE:i Ai!0NS 3/4.9.2 INSTRUMENTATION i

i 4 LIMITING CON 01"0N FOR OPERATION i

I 3.9.2.1 As a minimum, two trains of the Boron Oilution Mitigation System shall 4

I be OPERABLE ano operating with Shutdown Margin Alarm Ratios set at less than l

or eoual to 4 times the steady-state count rate, each with continuous indication j

in the control room.

j APPLICABILIT(:

MODE 6 l

t ACTION:

i (a) With one or both trains of the Boron Oilution Hitigation System

{

inoperaole or not operating, (I) immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes, and verify that valve NV-230 -

{

is closed ano secured within the next hour or (2) verify that two Source Range Neutron Flux Monitors are OPERABLE 4

and operating with Alarm Setpoints less than or equal to one-j nalf decade (square root of 10) above the steady-state count rate, each with continuous visual indication in the control i

room and one with audible indication in the control room and

?

c Rwfyrkkeap ue/rrrbp one with audible indication in the containment and verify that Ac& prese.Al la w the combined flowrate from ooth Reactor Makeup Water Pumps is i

C.re 9.rdly Ga R.p,t]ess than or equal toj0 gp within the next hour.

(b) Witt both trains of the Baron Oilution Mitigation System inoperable i

or not operating and 'one of-the Source Range Neutron Flux Monitors inoperable or not operating immediately suspend all operations l-involving core ALTERATIONS or positive reactivity changes and verify j

that valve NV-230 is closco and secured within the next hour.

4 (c) With both trains of the Beron Oilution Mitigation System inoperable or not operating and both of the Source Range Neutron Flux Monitors inoperable or not operating, determine the boron concentration of the Reactor Coolant System at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and verify that valve NV-230 is closed and secured within the next hour.

l SURVEILLANCE RE0VIREMENT 1

l 4.9.2.1.1 Each train of-the Boron Dilution Mitigation System.shall be demon-strated OPERABLE oy-performance of:

(a) A CHANNEL CHECK at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , (b) An ANALOG CHANNEL OPERATIONAt TEST within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to the initial start of CORE ALTERATIONS and' (c) An ANALOG CHANNEL OPERATIONAL-TEST-at least once per 31 days.

~

Amendment No.94(Un CATAWBA - UNITS 1 & 2 3/4 9k11 Amendment No.66(Unit 2)

l C - 14 i

J REFUEllNG OPE RTICNS SURVEILLANCE REOUIREMENTS (CONTINUED)

.:. 4 I

(d) At least once per 18 months the 80M5 shall be demonstrated OPERABLE oy:

(1) Verifying that each automatic valve actuated by the BDMS moves t:, its correct position upon receipt of a trip signal, and i

4

)

(2) verifying each reactor makeuo water pump stops, as designed.

. pen receipt of a trip signal, i

4.9.2.1.2 If using the Source Range Neutron Flux Monitors to meet the require-

]

ments of Tecnnical Specification 3.9.2. each Source Range Neutron Flux Monitor shall be cemonstrated OPERABLE by performance of:

1 (a) A CHANNEL CHECK at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , (b) An ANALOG CHANNEL OPERATIONAL TEST within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to the initial start of CORE ALTERATIONS or within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after declaring the BORON DILUTION MITIGATION SYSTEM inoperable, and (c) An ANALOG CHANNEL OPERATIONAL TEST at least once per 7 days.

l (d) The comoined flowrate from both Reactor Makeup Water Pumps shall be verified as less than or equal to # -gpe at least once per 7 days.

4ke. Reac% flakeup kisler t%f kwrsfe.

l prese M ;>, % Cm opud;9 ;~1r RpT L

l 4

1 i

l CATAWBA - UNITS 1 & 2 3/49-b" Amendment No.94 (Unit 1)

Amendment No.88 (Unit 2)

C - /5 REACTIVITY CONTROL SYSTEMS BASES BORATION SYSTEMS (Continued) j.;

MARGIN from expected operating conditions of 1.3% ak/k af ter xenon decay 1

Se and cooldown to 200'F.

The maximum expected boration capability requirement 3h occurs at EOL from full power equilibrium xenon conditions,and-requins--

oN 651-geWnt-of-7000 ppm-boeated vestee from the boric acid storage tanks 3 1 or 57,107 geWns-of-4000--ppm-boeated-wate" 'rce-the refueling water storage T.!

tank, s

',h[s The Technical Specification requires 42;400-geWas-of 7000 pp= berated.

g {

watee-from the boric acid tanks tc bc available-in Modes 1-4 This volume g 6-is based on the required volume for maintaining shutdown margin, unusable e e volume (to allow for a full suction pipe), instrument error, and additional op q (3 margin te ::ccunt-for -different-cores-and. conservatism as follows:

3 H,

preseded h & Care O erd;q

?-*

p j

Modes 1-4 Boric Acid Tank

.,l/s Sp.ct-

)3

"~ f Required volume for maintaining SDM

> - 0,051 gallcas--

ay K Additional Margin lW M -496 gallons 3*

Is Unusable Volume (to maintain full suction pipe) 7,230 gallons

g. 9 t &

n )

, Q7 14" of water equivalent n5 Vortexing (4" of water above top of suction pipe 2,066 gallons R g, a

,5ei Instrumentation Error (Based on Total Loop Acc.

1,550 gallons c

{jj for 1&2 NV5740 loops) - 2" of water equivalent 0

3 4 2.LvB3 gat 4cas q

a 3

-TMs-velee is increased-tc 22,000- gaWn: for ddi44cn 1 m:r9 m-d <'

4

{ }7 A similar approach is taken for calculating the required Refueling Water y

Storage Tank volume:

g.

h,E-

.2 2 When the temperature of one or more cold legs drops below 285 F in Mode 4,

-) {

the potential for low temperature overpressurization nf the reactor vessel g g-E.!

makes it necessary-to render one charging pump INOPERABLE and at least one

'j s

gj safety injection pump INOPERABLE.

The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to cr o verify all charging pumps except the required OPERABLE pump to be inoperable "j

below 285*F provides assurance that a mass addition pressure transient can P

be relieved by the operation of a single PORV.

),

V '

Refueling Water Storage Tank Requirements For Maintaining SDM - Modes 1-4 E-J Required Volume for Maintaining SDM o benhJia44y 4 A Ryd + --57,107 galica Cert 4*

r Unusable Volume (below nozzle) 13,442 gallons Instrument Inaccuracy 11,307 gallons Vortexing

_13,247 callons_

SkrW3 gaHbas-Addd.lonal M uss 3504 yil.ns

-The-Techn4 cal %ecificat4ca Volume 202,512 gs11on3 ss-determined by-

-eorreet4cq-thed.onk ' : low 1cve4-setpo4*t--(4 eve 4-et which makeup-4e-added-to-CATAWBA - UNITS 1&2 8 3/4 Ikg5 Amendment No. 82(Unit 1)

Amendment No. 76(Unit 2)

t C -kp REACTIVITY CONTROL SYSTEMS g

BASES BORATION SYSTEMS (Continued) i tanktfor-4nstrument-4naccuracy.

T h i+-lev el-p eev idefr4 he-ma x 4 mum-e va (4 a b l e

-volume-to-eceount-for-shutdown-eargio -woes-t-case-s4ngle-fe44ure -adequate-r r

-c o n ta inmen t-+ ump-volume-fo r-t ra n sfe e-to-ces t-ec ula t4 ony-a nd-su f14 ef en t-v olume--

e bove-t he-swi tc hove e-4nitt e t4 on-level-suc h-tha t-no-ope ra t o p-a c ti o n-4 s-requ ired-tefoe-to-ten-mi nu te s-a fter-the-4 nttia t4on-o f-the-aco4 den t, i

l With the coolant temperature below 200*F, one Boron Injection flow path is l

acceptable without single failure conrideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting j}j CORE ALTERATIONS and positive reactivity changes in the event the single Boron

/

njection flow path becomes inoperable, d W The boron capability required below 200*F is sufficient to provide a w t

),n SHU QOWN MARGIN of 1% ak/k after xenon decay and cooldown from 200'F to N

140'F. A--Thi+-condi-t4cn-requires-e4thee-685-gaWns-of-7000-ppm-borated-water

'}4 s

from the boric acid storage tanks or 3500-gaWns-of-2000-ppa berated-water-u 4pom-the refueling water storage tank.

e s

e E]Q The Boric Acid Tank and Refueling Water Storage Tank volumes required in Modes 5-6 to provide necessary SOM are based on the following inputs j

as discussed previously:

p,g,4 ;, q g Il Boric Acid Tank O *b.3 U "'Y' #P'

/

Ev Required Volume for maintaining SOM

-585-9a m ns-t Unusable Volume, Vortexing, Inst. Error 10

@ O; h additional margin 56 9 - -+,846 gallons

- & gallons _

l,y

.1.LA64 ga m as 34 Thi: value it inopeased-te-the Techn4ce4-Spec 444 eat 4on-value-of-14 000-7 2 ajahns-for-add 444ona4--mae9 nr 4

4, ycm.M i. fle, Care.

P 3 U~,% Rervt O rd y

Refueling Water Storage Tank 4)5 Required Volume for Maintaining SDM A -3dOO-gaWns A

Water Below the Nozzle 13,442 gallons Ey Instrument Inaccuracy 11,307 gallons Vortexing

_i3,247 gallons _

i e

2s E 496mys4 h s 7.*

/ld d,%.on J Mn ry.

3 508] p ilen r n

)}2

-Th44-value-4s-4cereased-te-the Techn4441-4pec44(cation-ve4ce-of d 5,000-ga W n: for add 444ona4-margin, o

f*

jj The contained water volume limits include allowance for water not available

.j because of discharge line location and other physical characteristics.

t o

>* A CATAWBA - UNITS 1&2 B 3/4 1-3316 Amendment No. 82 (Unit 1)

Amendment No. 76 (Unit 2)

I C.-17 i

l i

e t

REACTIVITY CONTROL SYSTEMS BASES 1

BORATION SYSTEMS (Continued) a

7. 5 an d 9.5" The limits on contained water volume and boron concentration of the refueling water storage tank also ensure a pH value of between 4.ai^

l for the solution recirculated within containment after a LOCA.

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

i The OPERABILITY of one Boron Injection System during REFUELING ensures l

that this system is available for reactivity control while in MODE 6.

~

i 3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that:

(1) acceptable power distribution limits are maintained. (2) the minimum SHUTOOWN MARGIN is main-tained, and (3) the potential effects of rod misalignment on associated accident.

analyses are limited.

OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits.

Verification that the Digital Rod Position Indicator agrees with the demanded position within i 12 steps 4

l at 24, 48, and 120 steps and fully withdrawn (> 225 steps) for the Control Banks and 18 and 210 steps and fully withdrawn for the Shutdown Banks provides assurances that the Digital Rod Position Indicator is operating correctly over the full range of indication.

Since the Digital Rod Position System does not i

indicate the actual shutdown rod position between 18 steps and 210 steps, only points in the indicated ranges are picked for verification of agreement with i

demanded position.

i i

l-3 i

CATAWBA - UNITS 1&2 B 3/4 1-ibl7 Amendment No,82 (Unit 1)

Amendment ho. 76 (Unit 2)

4 C - 18 i

1 l

3/4.5 EMERGENCY CORE COOLING SYSTEMS P

i BASES

}

l 3/4.5.1 ACCUMULATORS i

i The OPERABILITY of each Reactor Coolant System accumulator ensures that a sufficient volume of borated water will be immediately forced into the reactor core through each of the cold legs from the cold leg injection l

accumulators and-direct 4y inte the r.sacter vessa! 'r^- the epter heed %jection J

" elater-s in the event the Reactor Coolant System pressure falls below the pressure of the accumulators.

This initial surge of water into the core provides j

the initial cooling mechanism during large pipe ruptures, i

The limits on accumulator volume, boron concentration and pressure ensure j

that the assumptions used for accumulator injection in the safety analysis are

$rstert' **

I----?

The-a14ewed-down-44me-fer the :ce-elators-are variable be:ed epe-berer ---

l

-eeneentr:tien ta ensure:that the-reacter-i: thetd:rn-fe44ewig-a-LOCA and-thatr-j

-eny-prob)::: Or: corrected-in-a-44mely-manner.

Seberiticality i: :::er:d when l

-" "^-

-boren-concentret4en-4: eeve 1900 ppe, :: :dditional " -

'4-- l

-conc en t ra t4en-i s-abo ve-1800-ppm.

:encontrat4en-of-4::: than 1000 ppe ' :ny -

l

+(gle-accumulaton-on-a :--veleet weighted ever:;: ::y b: indicativ: ef :-pr:-

}

-blem,- ;uch-as-valve-leakage, but sin : -44ctor thetd:Ur 's-assured, additional--

-t4me-44-a44 r:d te-restore-boron-concentra4(en-4n-th :: emulators.

2 5

The accuc.ulator power operated isolation valves are considered to be

" operating eypasses" in the context of IEEE Std. 279-1971, which requires that j

bypasses of a protective function be removed automatically whenever permissive conditions are not met.

In addition, as these accumulator isolation valves fail to meet single failure criteria, removal of power to the valves is required.

l The limits for operation with an accumulator inoperable for any reason i

except an isolation valve ;1osed minimizes the time exposure of the plant to a LOCA event occurring concurrent with failure of an additional accumulator j

which 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 l

reactor in a mode where this capability is not required.

i i

CATAWBA - UNITS 1 & 2 8 3/4 5-1 Amendment No.101 (Unit 1)

Amendment No. 95 (Unit 2) pg y-e-w--t-

-r---wer--t-w=e--

r er

'-r-

-w e--ei e

--w w w w gvr-c e

---*w r

++4 e--e*--

v

-v-ev

-r-P--v-----'*r--e e--Te

="*

l C. lG i

linsertI A

The allowed outage time for the accumulators are variable based upon boron concentration to ensure that the reactor is shut down following a LOCA and that any problems are corrected in a timely manner. The minimum boron concentration required to ensure post LOCA suberiticality, as presented in the Core Operating Limits Itepon,is based on nominal accumulator volume conditions and allows additional outage time since 4

suberiticality is assured when the boron concentration is above this value. A slightly i

higher bomn concentration, the minimum accumulator boron concentration limit for LCO l

3.5.le presented in the Core Operating Limits lleport, is based on worst case liquid mass, i

boron concentration and meanrement errors. A concentration less than this LCO value in any single accumulator or as a volume weighted average may be indicative of a problem, such as valve leakage. Since reactor shutdown is assured if the boron concentration is above the minimum concentration to ensure post LOCA suberiticality and the accumulator volume is greater than or equal to the nominal volume, additional time is allowed to restore boron concentration in the accumulators.

i d

R 2-19

i C.Ao EMERGENCY CORE COOLING SYSTEMS j

BASES REFUELING WATER STORAGE TANK (Continued) 1 The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.

%5and1.j{)

The limits on contained water volume and boron concentration of the refueling water storage tank also ensure a pH value of between Br0-end-4r0-for the solution recirculated within containment after a LOCA.

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

The Technical Specification Volume 363,513 gallons was determinea by correcting the tank's low level setpoint (level at which makeup is added to tank) for instrument inaccuracy.

This level provides the maximum available volume to account for shutdown margin, worst case single f ailure, adequate containment sump volume for transfer to recirculation, and sufficient volume-above the switchover initiation level such that no operator action is required prior to ten minutes after the initiation of the accident.

2-20 CATAWBA - UNITS 1 & 2 B 3/4 5-3 Amendment No. 32(Unit 1)

Amenoment No. 23(Unit 2)

C-2)

\\

Changes to Core Operating Limits Report (Catawba 1 Cycle 7 & Catawba 2 Cycle 6) 2-21

\\

C 22 1

CNE14Wo.14 Page i of is Rev.m u

[

i i

a ie t

d l

Catawha Nuclear Station COLR i

l i

e r

4 I

i i

1 i-G j

Catawba Unit 2 Cycle /

Core Operating Limits Report-i 4

November 22,1991 i

i-t i

Duke Power Company l

Prepared by: M P S. O~k 2

Checked by: 'Mrc.

'NS$ )

4 4

v Approved by: [ A GSI-__

6 i

4

.T

-2 22

.,._.-u_.

C-23 4

CNEl W(x) 14 Parc 2 of 15 i

b Rev. tHxl j

Catawlia 2 C,scleg Core Operating 1.imits lleport i

)

REYlSION LOG 1

j Revnion Effective Date Effective Paces j

1

]

Original issue 22 November 1991 Pages 1 15 r

)

i a

i i

i a

t i

a 2

2-23 t

4

-.,,c

.v.--

,__w-.-._=,

r,__,

,.......r.

_ = -

,.mww

C 2#_. _ _

CNEl.tum.14 Pace 3 of 15 Cataw ha 2 Cy cle k(p Rev. (H K)

Core Operating Limits Report i

i i

i l.0 Core Ooerntine 1 imits Reoort This Core Operating Limits Report. (COLR). for Catawba Unit 2. Cycle has been prepared in accordance with the requirements of Technical Specification j

6.9.1.9.

The Tecnnical Specifications atfccted by this report are listed below:

.3 Moderator Temperature Coefficient 3/4.1.3..

'hutdown Rod insepmi j,-

j ge0 uc,4 1

3/4.1.3.6 ContfMbLinsertion Limit 43T6uxDifferg I"'rI I 3 /.t M..

Heat Flux Hot Channeih ir 3/4.2.3 Reactor Coolant System Flow id Nuclear Enthalpy Rise Hot Channel F2cin I

f 1

J 2 24

c ') s l

+

i l

insert i 1

2.2.1 Reactor Trip System Instrumentation Setpoints 3/4.1.1.3 Malerator Temperature Coefficient 3/41.2.5 llorated Water Source - Shutdown i

3/4.1.2.6 llorated Water Source Operating i

3/4.1.3.5 Shutdown Rod Insertion Limit 3/4.1.3.6 Control Rod Insertion Limit 3/4.2.1 Axial Flux Difference 3/4.2.2 11 eat Flux flot Channel Factor i

3/4.2.3 Nuclear Enthalpy Rise flot Channel Factor j

3/4.3.3.12 lloron Dilution Mitigation System 3/4.5.1 Accumulators

{

3/4.5.4 Refueling Water Storage Tank i

3/3.9.2 Instrumentation 1

l r

i

)

l J

l 2-25

.=.

I O Mh i

CNEl.ou o.14 l' age 4 of i$

Rev.tHe Cataulia 2 C cle/ Core Operating 1.itnits Report 3

6 d.i

,KO Oceratine 1.imits i

R: cycle specilie parameter limits for the specifications listed in section 1.0 are presented in the following subsections. These limits have been developed using

.NRC.apprmed methodologies specified in Technical Specification 6.9.1.9.

Liseef 2 --+

E!' Moderntor Temocrature Coefficient 60ccificatlun 3/4.1.1.3) 30 I

M.1 The Moderator Temperature Coefficient (MTC) Limits are:

Ja The \\1TC shall be less positive than the limits shown in Figure 1. The BOC/ARO/HZP MTC shall be less positive that 0.7

10 d AK/K/F.

l The EOC/ARO/RTP MTC shall be less negative that -4.1

10-d aK/K/ F.

,26.2 The MTC Surveillance Limit is:

30 The 300 PPM /ARO/RTP MTC should be less negative than or equal to i

-3.2

10 4 AK/K/~F.

Where:

BOC stands for Beginning of Cycle ARO stands for All Rods Out HZP stands for Hot Zero (Thermah Power EOC stands for End of Cycle RTP stands for Rated Thermal Power 2 26 t

c. a1 4

i i

insert 2 2.0 iteactor Trin Ssstem 1938rumen, talion Setuoints (Soccification 2.2.1) 1 2.1 Overtemperature AT Setpoint Parameter Values J

4 Parameter Value K = 1.1953 Overtemperature AT reactor trip setpoint 1

1 K = 0.03163/ F Overtemperature AT reactor trip heatup setpoint 2

j penalty coefficient K = 0.001414/ psi I

Overtemperature AT reactor trip depressurization 3

.setpoint penalty coefficient Measured reactor vessel ATlead/ lag time tt = 12 sec.,

j constants 12 = 3 sec.

)

Measured ATlag time constant t3 = 0 sec.

Measured reactor vessel average temperature ta = 22 sec.

d lead / lag time constants 15 = 4 see.

Measured reactor vessel average temperature lag T6 = 0 sec.

i time constant 1

J f (AI) "posi'ive" breakpoint

= 3.0% Al i

f g(AI) " negative" breakpoint

= 39.9% Al f (AI)" positive" slope

= 2.316% ATo / WAl i

f (AI)" negative" siope

= 3.910% ATo / 7eAl t

2 27

.. = -

b 0-0E 4

I Insert 2 - continued 2.2 Overpower AT Setpoint Parameter Values 7

j Parameter Value i

K = 1.0819

}

Overpower AT reactor trip setpoint 4

4 K = 0.001291/ F Overpower AT reactor trip heatup setpoint penalty

,5 6

coefficient l

Measured reactor vessel ATlead/ lag time t g = 12 sec.,

I constants 12 = 3 sec.

i i

Measured ATlag time constant 13 " O Sec-i i

Measured reactor vessel average temperature lag t6 = 0 sec.

time constant Measured reactor vessel average temperature rate-t7 = 10 sec.

lag time constant 3

f (Al)" positive" breakpoint

= 35.07c Al 2

f (AI)" negative" breakpoint

= -35.07c Al 2

l f (AI)" positive" slope

= 7.07c ATo / 7 cal 2

f (Al)" negative" slope

= 7.07c ATn / 7 cal 2

i 2-28

c.A9 J

CNCI m0014 Page5ogI5 Rev.iul i

Catawba 2 Cycle

' ore Operating Llanits Report t

I C U.9 --

4 x 0.8 -

1 j

Unacceptable Operation j 0.7

=

j 0.6 --

'oj Acceptable Operauon 8

E3 0.4 ---

Em Q 0.3 --

s-U.2 ~*

b 1

?:= 0.1 0

0 10 20 30 40 50 60 70 80 90 100 Percent of Itated Thermal Power Figure 1 Wderator Temperature Coefficient Versus Power Level 2-29

~

Q 30 CNiil miWi 14 Page n 01 15 Rev,001 Catau ha 2 C,s cle/ Core Operating Limits 1(eport G

ZnserT3 ->

~

/ff Shutdou n flod Insertion 1.ifnit (Suecification 3'J.l.3.!)

3O 222 Jer.1 The shutdown rods shall be withdrawn to at leastJEteps.

33 Jrf L'ontrol itud insertion 1.ltnits tSoccification 3/4.1.3.6) 34 26.1 Tne control red banks shall be limited in pnysical insemon as shown in 38 Figuie 2.

M hmi I:los I)lfference doecification 3/4.2.1) 3.S y

1 - 'u \\X!-il M ' O!FFE"CNCE i AFD, Lanin am go-pis Etgun p

%.2 The target ti during base load opersti n is not applicable for 3J Catawba 2 C:cic /s

)d,p mnimum all wable pow th peration ( AFLND)

Ac w appii--

Jawen uyucg#

s

)di Re Axial Flux Difference (AFD) Limits are provided in Figure 3.

3'5 COLR

( AFD Limit)negauve is the negative AFD limit from Figure 3.

( AFD Limit)COLR is the positive AFD lindt from Figure 3.

po,n;y, 4

l l

l i

2 30 h

C-3I Insert 3 3.1 llorated Water Source. Simidown (Soccification 3/4,1.2.5 )

J 3.1.1 Volume and boron concentrations for the lloric Acid Storage System and the l

Refueling Water Storage Tank (RWST)during modes: 5 & 6:

Parameter Limit l

Boric Acid Storage System minimum boron 7,000 ppm concentration for LCO 3.1.2.5a

{

Boric Acid Storage System minimum contained 12,000 gallons water volume for LCO 3.1.2.Sa Borie Acid Storage System minimum water volume 585 gallons required to m.tintain SDM at 7,000 ppm Refueling Water Storage Tank minimum boron 2,000 ppm concentration for LCO 3.1.2.5b Refueling Water Storage Tank minimum contained 45.000 gallons water volume for LCO 3.1.2.5b Refueling Water Storage Tank minimum water 3,500 gallons volume required to maintain SDM at 2,000 ppm 3.2 llorated Water Source - Operatine (Suecification 3/4.1.2.6 )

3.2.1 Volume and boron concentrations for the Boric Acid Storage System and the Refueling Water Storage Tank (RWST) during modes: 1,2,3 & 4:

Parameter Lignit Boric Acid Storace System minimum boron 7,000 ppm concentration for LCO 3.1.2,6a Borie Acid Storage System minimum contained 22,000 gallons water volume for LCO 3.1.2.6a Borie Acid Storage System minimum water volume 9,851 gallons required to maintain SDhl at 7,000 ppm 2-31

~

C 32 4

I Insert 3 - continued Parameter Limu Refueling Water Storage Tank minimum boron 2,000 ppm concentration for LCO 3.1.2.6b Refueling Water Storage Tank minimum contained 98,607 gallons water volume for LCO 3.1.2.6b Refueling Water Storage Tank minimum water

$7,107 gallons volume required to maintain SDM at 2,000 ppm i

4 i

I l

i 2-32

(~33 CNE!450014 i

Page 7 of 15 Rey, tXX) i 1

Catawba 2 Cycle 5 Core Operating Limits Report 4

Fully Witherawn \\

(27.4 %,226)

(77.8 % 226) l R c e u e_.

i 220 larYk i

~

200

^5 "Y BANK i

~

180 (0%.163)

(100 % 161)

~"

'50 E

2 40

~

3 BA C

=

3

~

20 i

3;-

b 100

~

5 1

3 30 f

8 NK D.

4

~

3 50 3

(0%.47)

~

40

~

20 (30%0)

Fully inserted

/!

I I

1\\

l i

I I

O O

20 40 60 80 100 Relative Power (Percr.t)

Figure 2 4

Control Rod Bank insertion Limits Versus Percent Rated Thermal Power 2-33 n'e r

l C 3y 4

e i

1 e

ie i

j

-Insert 4 j

Catawba 2 Cycle 6 Core Operating Limits Report i

4 i

1 Fully Withdrawn (29 E 230) (Maximum =230)h (80E 230) 230

~ ~ ~ ~ " - - - - ~ ~ - ~ ~ ~ ~~~~ ~ ~ ~ ~ ~

"'g'- Fully Withdrawn

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

"O -

4 i

i 200 --

(Minimum =222) g I

IIA N K 11 180 - -

E i

2 (100 % 161) l j 160 --(OLty) s e

i g.140--

I d-IIANK C a 120 --

O a

j f100--

i 8

j 80 -

IIANK D

,B.

4 l

g 60 --

4 m

J l

40 --(0148) i 20 --

Fully inserted (30p,0)

O i

i.

i i

~

0-10 20 30 40 50-60 70 80 90 100 i

Percent of Rated Thermal Power j

Figure 2 -

4-Control Rod Bank Insertion Limits Versus Percent of Rated Thermal Power 4

s I

4 s

t 3-2-34 5

.,,..,......,,J

,, _,. ~ ~ -

'C-35 CNEl-(M0014 j

Page 8 of 15 Rev.000 I

i-i t

i

?

4 h

)

Catawba 2 Cycle k Cure Operating Limits Report i

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Figure 3 Percent of Rated Thermal Power Versus Axial Flux Difference Limits l

i

-l i

s 2-35

0. ~ 3&

CNEl N00-14 Page 9 of 15 Rev.000 Catawba 2 Cycle 5 Core Operating Limits Report 2.5 lies Flux flot ('hnnnel Factor. F (Z) (Specification 3/4.2.2)

Q F (Z) c. ""

Kf Z) for P > 0.5 9

F (Z) s F,"

Kt )

for P s 0.5 9

0.5 Dermal Pos r where: P=

Rated Thermal Po er 2.5.1 F"" = 2.32 0

2.5.2 K(Z)is provided in Figur 4.

2.5.3 W(Z) values are prov' ed in Figures 5 thro 'h 7.

2.5.4 Base load W(Z)' are not applicable for Catawba ' Cycle 5.

I

/

2-36

-- -... -..-. -. - ~

G -3'1 l

4 i

i i

insert 5

- Catnwba 2 Cycle 6 Core Operating Limits Report i

A.

i 3.6 llent Flux Ilot Channel Factor.FQ(X.Y.Z)(Snecifiention 3/4.2.2) l RTP 3.6.1 F

= 2.32 9

i 3.6.2 K(Z) is provided in Figure 4 for Mark-BW fuel.

I 3.6.3 K(Z) is provided in Figure 5 for OFA fuel.

l The following parameters are required for core monitoring per the Surveillance j

Requirements of Specification 3/4.2.2:

F (X,Y,Z)

M (X,Y,Z)

Q 3.6.4 (F (X,Y,Z)]OP =

where IF (X,Y,ZilOP=

cycle dependent maximum allowable design peaking j

factor which ensures that the F (X,Y,Z) limit will be Q

j p served for operation within the LCO limits.

(F (X,Y,Z)]OP ncludes allowances forcalculational and l

i measurement uncertainties.

D F (X,Y,Z) = the design power distribution for F. F (X,Y,Z) is provided g

Q j

in Table 2 for normal operation and tabl 2A for power

)

escalation testing during initial startup, 2

l MQ(X,Y,Z) = the margin remaining in core location X,Y,Z to the LOCA -

limit in the transient power distribution. M (X,Y,Z) is Q

. provided in Table 3 for normal operation and table 3 A for power escalation testing during initial startup.

4 UMT = Measurement Uncertainty (UMT = 1.05).

MT = Engineering Hot Channel Factor (MT = 1.03).

l TILT = Peaking penalty that accounts for allowable quadrant power tilt ratio of1.02.

i

[F (X,Y,Z))OP s the parameter identified as FMAX(X,Y,Z)in DPC-NE-L i

NOTE:

9 q

3 20llPA.

i I

i 2,

1 v

e

-e,

,.,s.

-n--,,

C - 38 Insert 5 - continued Catanha 2 Cycle 6 Core Operating Limits Report j

D li f(X,Y,Z))RPS = O(X Y Z)

M (X,Y,Z)

F C

g 3.6.5 where iF (X,Y,Z)]RPS =

cycle dependent maximum allowable design peaking factor which ensures that the centerli fuel melt limit will be preserved for all operation. [F (X,Y,Z)]RPS includes allowac.ces for calculational d measurement uncertainties.

D D

F (X,Y,Z) = the design power distributions for F. F (X,Y,Z)is provided 9

Q 9 in Table 2 for normal operation and table 2A for power escalation testing during initial startup.

i M (X,Y,7) = the margin remaining to the CFM limit in core location X,Y,Z C

from the transient power distribution. M (X,Y,Z)

C calculations parallel the M (X,Y,Z) calculations described in Q

DPC-NE-201 IPA, except that the LOCA limit is replaced with the CFM limit. M (X,Y,Z) is provided in Table 4 for C

normal operation and table 4A for power escalation testing during initial startup.

UMT = Measurement Uncertainty (UMT = 1.05).

MT = Engineering Hot Channel Factor (MT = 1.03).

i TILT =. Peaking penalty that accounts for allowable quadrant power tilt ratio

[

of1.02.

Mp NOTE: lF (X,Y,Z)]RPS is similar to the parameter identified as FM AX(X,Y,Z) in L

k DPC NE-2 1 IPA except that M (X,Y,Z) replaces Mg(X,Y,Z).

C 3.6.6 KSLOPE = 0.078 where KSLOPE = Adjustment to the K valu fiom OTAT required to compensate 1

for each 1% that (F (X,Y,Z)]RPS exceeds it limit 1

2-38

)

l-

,c-31 CNEl 4

Page 10 15 R, txx) 4 Y

'atawba 2 Cycle 5 Core Operating Limits eport 1

I i

l 1.2

/

) (0.0,1.00) ;

l(6.[ 1.00)I g(10.8.o.94) 1 I

O,8 1l 8

l

@0.6 (12.0.0.647) 1 0.4 i

0.2 l

]

0 O

2 3

4 5

6 7-8 10 11.12 i

Core Height (ft).

Figure 4 i

K(Z), Normalized F (Z) as a Function of Core Height Q

kefacA wik 1 5erf b_

l

/

n 2 39

C.-40 Insert 6 Catawba 2 Cycle 6 Core Operating Limits Report 1.2

)(0.0.1.00)

(8.0,1.00)

I (10.8.0.94) 0.8 g0.6 (12.0.0.647)l 0.4 0.2 0

0 1

2 3

4 5

6 7

8 9

10 11 12 l

Core lleight (f0 l

Figure 4 K(Z), Normalized F (X,Y,Z) as a Function of Core Height for MkBW Fuel Q

l 2-40

C-4/

Insert 6 - continued Cataw tia 2 Cycle 6 Core Operating I imits Report 1.2 (0.0.1.00)

(6.0,1.00) 1 (10.8.0.94) 0.8 h0.6 (12.0.0.647) 0.4 0.2 I

f 9

0 1

2 3

4 5

6 7

8 9

10 11 12 Core lleight (ft)

Figure 5 K(Z), Normalized F (X Y,Z) as a Function of Core Height for OFA Fuel Q

2-41 J

O-Y1 CNEl 040014 Pace 11 of i Rey, t Catawba 2 Cycle 5 Core Operating Limits Report E! Cit?

BOL trt m wm 1.55. x

0.00 1.0000 0.20 1.0000

' 6

\\

i * *

0.40 1.0000 s

i i i e

i e

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/

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i i.

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i

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1.40 1.0000 N

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e i e i i ofi e 1.80 1.4776 A

. i i i i e i i e i i i j i i i

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8 10 12 bon::a CORE HEIGHT (rIIT) gep Figure 5 Catawba Unit 2 Cycle 5 hh(___

RAOC W(Z) at ig1WD/MTU

/

Lp and Bottom 151 excluded as per Tech. Spec. 4.2.2.2.G

C-43 CNEl-0400- 14 /

Page 12 of 15 Rev. o(

Catawba 2 Cycle 5 Core Operating Limits Report KEIClif MOL (rEET)

W(1)

0.00 1.0000 1.55. \\

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8 10 12 bestem CCRE HEIGHT (TEET) top

/

Figure 6 Catawba Unit 2 Cycle 5 RAOC WIZi at g\\ LWD /MTU bdEff s

10.md Bonom i50 emudeu as per Tech. 50cc..2.2.2.G

[y c-q#

3, e

l CNElO m I4 j

Page 13 of 15 Rev,in 1

l Catawba 2 Cycle 5 Core Operating Limits Report l

4 I

MI!Olt?

F.0L (TIIT) 18(El Ik O.00 1.0000

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s i

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2 4

6 8

10 12 tott:m CCE HEIGRT (FEET) top Ficure 7 b6 Catawba nit 2 Cvele 5

/

RAOC WIZ) at 13000 $1WDiMTU 2-44 Ton and Bottom iM excluded as per Tech. Spec. C.2.2.0

]

~ - - -.-

i CNEl GWM4 /

Pace 14 of 1 Rev.t# L j

Catawba 2 Cy cle 5 Core Operating Limits Report i

4 2.6 CC Flow Rate and Nucienr Enthniny Rise Hot Chnnnel Factor. N,g (Sp\\eqificatitm 3/4.2.3)

R=

l pRTP,}, ( [ + M p6g

([.p))

a d

i j

Therm Power j

where: P=-

j -

Rated The al Power i

j-2.6.1 FRTPAH " I*O j

2.6.2 MF g = 0.3 3

l 2.6.3 The Acceptable Operat' n Regt n from the combination of Reactor Coolant System total ow and R provided in Figure 8.

i f

i

/

I

/

i i

i J

a i-t 9

eF O

J 2-45 a

v

n

I 4

i i

insert 7 Catawba 2 Cycle 6 Core Operating Limits Report i

i l

j 3,7 Nuclear Enthalov Rise 1101 Channel Factor. F ggiX.Y.21(Snecification 3/4.2.3) 3 l

1 The following parameters are required for core monitoring per the LCO Requirements of l

SpeciGcation 3/4.2.3:

)

l 3.7.1 (Fag g(X,Y)]LCO = M ARP (X,Y)

_l 0 + RRI-l * (1.0 - P) j where (M ARP(X,Y)) = Catawba 2 Cycle 6 Operating Limit Maximum Allowable Radial Peaks. (MARP(X,Y))is provided in Table 1.

Thenml Power e

p

Rated Themial Power The following parameters. are required for core monitoring per _the Surveillance i

Requirements of Specification 3/4.2.3:

l

-D F

(X,Y)

Mag (X,Y) g AH(X'Y)]SURV " 3g 3*7*'

lF UMR

TILT

~

l' I

where IFfg(X,Y)lSURV =

cycle dependent maximum allowable design peakir.g factor which ensures that the F )1(X,Y) limit will be A

pr(served for operation within the LCO limits.

AH(X,Y)]SURV ncludes allowances for calculational l

[F i

j and measurement uncertainties.

i

-- D D

F3g(X,Y) = the design power distribution for FAH.FAH(X,Y)is provided in Table 5 for normal operation and table 5A for power escalation testing during initial startup.

i L

May(X,Y) = the margin remaining in core location X,Y to the Operational -

l DNB limit in the transient power distribution. MAH(X,Y) is provided in Table 6 for normal operation and table 6A for power escalation testing during initial startup.

i UMR = Uncertainty value for measured radial peaks,(UMR = 1.04).-

TILT =_ Peaking penalty that accounts for allowable quadrant power tilt ratio i

of1.02.

i 2-46

.__._..._._q C 41 i

Insert 7 - continued Catawba 2 Cycle 6 Core Operating Limits Report 4

L MAX SURV (X,Y) in DPC-NE-t NOTE: (F3g(X,Y)l is the parameter identified as FAH t-201 IPA.

l 3.7.3 RRH = 3.34 where RRil = Thermal Power reduction required to compensate for each 1% that j

Fay (X,Y) exceeds its limit.

3.7.4 TRH = 0.04 where TRH = Reduction in OTAT K setpoint required to compensate for each 1%

1 that FaglX,Y) exceeds its limit.

i i

3.8 Baron 1)ilution Mitiention System (Snecification 3/4.3.3.12)

)

[i 3.8.1 Reactor Water Makeup Pump flowrate limits:

Aonlicable Mode Linul Mode 3 or 4 s 150 gpm Mode 5 6 70 gpm 3.9 Accumulators (Soccification 3/4.5.1) 3.9.1 Boron concentration limits during modes: 1,2 and 3;

- Parameter Limits Cold Leg Accumulator minimum baron 1,900 ppm concentration for LCO 3.5.lc Cold Leg Accumulator maximum baron 2,100 pp_m concentration for LCO 3.5.lc 2-47

)

E-W 4

i i

i Insert 7 - continued l

Catauha 2 Cycle 6 Core Operating Limits Report 4

Minimum Cold Leg Accumulator boron 1,800 ppm i

concentration required to ensure post-LOCA l

subesiticality a

3.10 Refueline Water Storace Tank (Soccification 3/4.5.4) i e

3.10.1 Boron concentration limits during modes: 1,2,3 and 4:

j Parameter Limits l

Refueling Water Storage Tank minimum boron 2,000 ppm i

concentration for LCO 3.5.4b i

i Refueling Water Storage Tank maximum boron 2,100 ppm j'

concentration for LCO 3.5.4b -

i 3.11 Instrumentation (Snecification 3/4.9.2) j 3.11.1 Reactor hiakeup Water Pump Flowrate Limit:

Aonlicable Mxle Limil N1 ode 6 s70gpm-1 1

1 i

i l

l 2-48 4-

E G-@

1 i

Lo insert 7 - contilmed i

Catawba 2 Cycle 6 Core Operating Limits Report i

Table 1.- Maximum Allowable Radial Peaks (MARPs) 1

)

13 Axial Peak 1.4 Axial Peak j

Core lleicht 1.1 Axial Peak 1.2 Axial Peak -

U0 MARP MARP MARP MARP i

0.12-

-1.5809 1.6266-1.6722 1.7113 1.2 1.5806 1.6259 1.6677 1.7085 f"

2.4 1.5836 1.6265 1.6663 1.7025 3.6 1.5859 1.6263 1.6635 1.6960 4.8 1.5871 1.6240 1.6571 1.675I i-6.0 1.5878 1.6196 1.6470 1.6303 l

7.2 1.5864 1.6130 1.6265 1.5848

)

8.4 1.5781 1.5956-1.5773 1.5327 9.6 1.5655 1.5612 1.5208 1.4815 10.8 1.5459 1.5152-1.4717 1.4292 g

12.0 1.5133

.4693 1.4274 1.3878 l

Core Heicht 1.5 Axial Peak 1.6 Axial Peak 1.7 Axial Peak 1.8 Axial Peak l

dQ MARP MARP MARP MARP 1

0.12 1.7477 1.7331 1.7054 1.6438 j

1.2 1.7433 1.7029 1.6789 1.6193 2.4 1.7126 1.6616 1.6433 1.5869 3.6 1.6735 1.6211

-1,6011 1.55(M 4.8 1.6313 1.5811

-1.5622 1.5121 l

6.0 1.5868

_l.5415

_l.5238 1.4763 i

7.2 1.5378 1.4913' t,4766 =

1.4344 8.4

' t.4886 1.4450 1.4296 1.3880 9.6 1.4399-1,4013 1.3882-1.3490 l.3526 1.3433 1.3081-j 10.8

_l.3883

_ 1,3140 1.3078 1.2749' l

12.0 1.3500 Core Heicht 1.9 Axial Peak 2.1 Axial Peak HQ MARP MARP 0.12 1.5839 1.5401 1.2

-1.5624-1.5154 -

l 2.4 1.5328 1.4801-3.6 1.5013-1.4395 -

l 4.8 1.4626 1.4030 6.0 1.4291 1.3619 7.2 1.3920 1.3271 I

8.4 1.3485 1.2824-i 9.6 1.3126--

1.2501 10.E 1.2726 1.2091 2-49

_ _ =

C-50 a

CNEl 040014f Pace 15 of 5 Rev 00 Catawba 2 Cscle 5 Cure Operating Limits Report N

Penaities of 0.1% or unestected fesowater ventun fouling and measutoment une aintro; of 2.1% for ficw ano 4 0% for incore' easurement of F'2Hare includ in mis 0gure.

l 40.0 39.5 rmissible p,,g g g g g,g Ob '* "

Operation Reg n Regton 39.0 -

g (1.000 38.760)

S Restncted Operation Region (Power < 98%

)

c.

38.5-

/

~ (0.994 38.372)

Restncteo Operation Region (Power i % RTP) w 1

i 28.0 (0.988, 37.985)

Restncted Operation Region (Powelr 194% RTP)

$o

/

~

(0.982, 37.597) l E 27.5 -

Restnctea 0::eration Regio (Power 192% RTP) 1

~

(0.977. 37.210)

[J 37.0 - Restnctec Ocoration gion (Power <90% RTP)

(0.971. 36.822) l i

I 36.5 -

E i

l 25.0 i

0.93 0.94 0.95 0.96 0.97 0.98 0.99 00 1.01 1.02 R FOUR LOOPS IN OPERATION beJA Figure 8 RCS Flow vs. R.Four 1.oups in Operation 2-50

C. - s i 1

1 1

CNEl-0400 24 (Rev. 0) 60f302 Catawba 1 Cycle 7 Core Operating Limits Report 1.0 Core Oneratine I imits Renort

'Ihis Core Operating Limits Report, (COLR), for Catawba, Unit 1, Cycle 7 has l

been prepared in accordance with the requirements of Technical Specification 6.9.1.9.

The Technical Specifications affected by this report are listed below:

i

}#fl A

SQ3 Moderator Temperature Co 3/4.1.3.

utdown Rod Insertion li f

f 3/4.1.3.6 Contr Ind. imit 3,g

{

4 3/4.2.1 Axial -

iffii 3/4.2.2 i cat Flux liot Channel i

3/4 Nuclear Enthalpy Rise Hot Channe I

4 I

i l

I I

4 2-51 4

4

m.,,

C - 52.

Insert 1 2.2.1 Reactor Trip System Instrumentation Setpoints 3/4.1.1.3 Moderator Temperature Caificient 3/4.1.2.5 Borated Water Source - Shutdown 3/4.1.2.6 Borated Water Source - Operating 3/4.1.3.5 Shutdown Rod Insertion Limit 3/4.1.3.6 Control Rod Insertion Limit 3/4.2.1 Axial Flux Difference 3/4.2.2 Ileat Flux Hot Channel Factor 3/4.2.3 Nuclear Enthalpy Rise Hot Channel Factor 3/4.3.3.12 Baron Dilution Mitigation System 3/4.5.1 Accumulators 3

3/4.5.4 Refueling Water Storage Tank 3/3.9.2 Instrumentation i

1 2 52

~

C 53

)

i s

CNE!4400-24 (Rev. 0) 7 of 302 j

Catawba 1 Cyle 7 Core Operating Limits Report i

^

d.L g.0 Ooeratino Limits The :ycle specific parameter limits for the specifications listed in section 1.0 are presented in the following subsections. These limits have been developed using NRC approved methodologies specified i' Technical Specification 6.9.1.9.

i-.T.s sc.t 2. --+

2:f Moderator Temocrature Coefficient (Soecification 3/4.1.1.31 30 M.1 The Moderator Temperature Coefficient (MTC) Limits are:

3.0 The MTC shall be less positive than the limits shown in Figure 1. The l

BOC/ARO/HZP MTC shall be less positive that 0.7

10-4 AK/K/ F.

The EOC/ARO/RTP MTC shall be less negative that -4.1* 10 4 AK/KfF.

,2:f.2 The MTC Survei; lance Limit is:

i 3.0 The 300 PPM /ARO/RTP MTC should be less negative than or equal to i

3.2

10 4 AK/KrF.

i Where:

.BOC stands for Beginning of Cycle ARO stands for All Rods Out HZP stands for Hot Zero (Thermal) Power i

EOC stands for End of Cycle'-

RTP stands for Rated Thermal Power -

t 7

k h

2-53 qry r-a y

e

~m

C.-54 Insert 2 2.0 Reactor Trio System Instrumentation Setooints (Snecincation 2.2.1) 2.1 Overtemperature AT Setpoint Parameter Values Parameter Value K = 1.1953 Overtemperature AT reactor trip setpoint 1

i K = 0.03163/ F Overtemperature AT reactor trip heatup setpoint 2

penalty coefficient K = 0.001414/ psi Overtemperature AT reactor trip depressurization 3

setpoint penalty coefficient

]

Measured reactor vessel ATlead/ lag time ti = 12 sec.,

constants t2 = 3 sec.

Measured ATlag time constant 13 = 0 sec.

Measured reactor vessel average temperature T4 = 22 sec.

lead / lag time constants T5 = 4 sec.

Measured reactor vessel average temperature lag T6 = 0 sec.

time constant i

f (AI)" positive" breakpoint

= 3.0% Al i

f (AI)" negative" breakpoint

= -39.9% Al i

f (AI) " positive" slope

= 2.316% AT / WAl 1

o f (AI)" negative" slope

= 3.910% AT / %AI 1

o 2-54

C-SS Insert 2 - contintied 2.2 Overpower AT Setpoint Parameter Values Parameter Value Overpower AT reactor trip setpoint K4 = 1.0819 K = 0.001291/ F Overpower AT reactor trip heatup setpoint penalty 6

coefficient Measured reactor vessel AT lead / lag time t1 = 12 sec.,

constants t2 = 3 sec.

Measured AT lag time constant 13 = 0 sec.

Mcasured reactor vessel average temperature lag T6 = 0 sec.

time constant Measured reactor vessel average temperature rate-17 = 10 sec.

lag time constant f (AI)" Positive" breakpoint

= 35.0% Al 2

f (AI)" negative" breakpoint

= -35.0% Al 2

f (Al)" positive" slope

= 7.0% AT / %AI 2

o f (AI)" negative" slope

= 7.0% ATo / %Al 2

l i

2-55 I

C-s4 CNEl 0400 24 (Rev. 0) 9 0f 302 Catawba 1 Cycle 7 Core Operating Limits Report Luc.cf 3 --9 i

.2d Shutdown Rod Insertion Limit (Snecification 3/4.1.3.5)

.33

,2:7.1 The shutdown rods shall be withdrawn to at least 222 steps.

~

33 p.5 Control Rod Insertion I imits (Snecification 3/4.1.3.6) 34

.26.1 The control rod banks shall be limited to physical insertion as shown in af Figure 2.

g Axini Flux Difference (Soecification 3/4.2.1) 3.5 J:(1 The Axial Flux Difference (AFD) Limits are provided in Figure 3.

36

( AFD Limit)COLRis the negative AFD limit from Figure 3.

3 negative

^

( AFD Limit)COLR is the positive AFD limit from Figure 3.

posniy, i

}

2 E

i i

4 d

i 2-56

Q.Sr1 i

i l

Insert 3 3.1 Horated Water Source Shutdown (Snecification 3/4.1.2.51 3.1.1 Volume and boron concentrations for the Boric Acid Storage System and the Refueling Water Storage Tank (RWST) during modes: 5 & 6:

l Parameter Limit Boric Acid Storage System minimum baron 7,000 ppm i

concentration for LCO 3.1.2.5a

(

Boric Acid Storage System minimum contained 12,000 gallons water volume for LCO 3.1.2.5a i

{

Boric Acid Storage System minimum water volume 585 gallons required to maintain SDM at 7,000 ppm i

Refueling Water Storage Tank minimum boron 2,000 ppm concentration for LCO 3.1.2.5b' Refueling Water Storage Tank minimum contained 45,000 gallons j

water volume for LCO 3.1.2.5b Refueling Water Storage Tank minimum water 3,500 gallons volume required to maintain SDM at 2,000 ppm l-3.2 Itorated Water Source - Ooeratine (Snecification 3/4.1.2.6 )

3.2.1 Volume and boron concentrations for the Boric Acid Storage System and the Refueling Water Storage Tank (RWST) during modes: 1,2,3 & 4:.

i-l

. Parameter Limit j

Boric Acid Storage System minimum boron 7,000 ppm -

concentration for LCO 3.1.2.6a l

Boric Acid Storage System minimum contained -

22,000 gallons water volume for LCO 3.1.2.6a Boric Acid Storage System minimum water volume = 9,851 gallons i

required to maintain SDM at 7,000 ppm i

4 2 57 1

-'-s---

e.-

e eryyr p pti1

-e-F q

eiw

,eg i._.,.a-yp-r-a 1g-+t--

-t7 3--PP'gr.+1?'=-rwt-4

- = =F ?!

C-se Insert 3 - continued Parameter LEntt Refueling Water Storage Tank minimum boron 2,000 ppm concentration for LCO 3.1.2.6b Refueling Water Storage Tank minimum contained 98,607 gallons water volume for LCO 3.1.2.6b s

Refueling Water Storage Tank minimum water 57,107 gallons volume required to maintain SDM at 2,000 ppm J

s i

4 1

2-58

c. 59 1

i CNEl.0400 24 (Rev. 0) 12 0f 302 Catawba 1 Cycle 7 Core Operating Limits Report 2f IIcat Flux Hot Channel Fnctorfq(X.Y.2)(Snecifiention 3/4.2.2) 3.(e Jhi.1 F"# = 2.32 A

3.6 x

l E!I.2 K(Z) is provided in Figure 4 for Mark-BW fuel.

3.G 2:5.3 K(Z)is provided in Figure 5 for OFA fuel.

34 The following parameters are required for core monitoring per the Surveillance j

Requirements of Specification 3/4.2.2:

D F (X,Y,Z)

Mg(X,Y,Z) q M4 [E(X,Y,Z)]OP =

g 34 l

where [E(X,Y,Z)]OP = cycle dependent maximum allowable design peaking n

factor which ensures that the F (X,Y,Z) limit will be Q

preserved for operation within the LCO limits.

[k(X,Y,Z)]OP includes allowances for calculational and g

measurement uncertainties.

Th(X,Y,Z) = the design power distribution for Fq. F (X,Y,Z) is provided D

g j

in Table 2 for normal operation and table 2A for power escalation testing during initial startup.

i Mn(X,Y,Z) = the margin remaining in core location X,,Z to the LOCA limit in the transient power distribution. Mg(X,Y,2) is provided in Table 3 for normal operation and table 3A for l

. power escalation testing during initial startup.

UMT = Measurement Uncertainty (UMT = 1.05).

i MT = Engineering Hot Channel Factor (MT = 1.03).

TILT = Peaking penalty that accounts for allowable quadrant power tilt ratio of 1.02.

-NOTE:-[E(X,Y,Z)]OP is the parameter identified as F (X,Y,Z) in DPC-NE-g 201 IPA.

2-59 2

4 e

..e

~

,-e.,

a 4

C-60 CNE! 0100 24 (Rev. 0) 13 of 302 Catan ha 1 Cycle 7 Core Operating Limits lleport f((X,Y,Z)

M (X,Y,Z)

C

,25.5 [I (X,Y,Z))RPS "

UMT

MT

TILT g

3.b where [IE(X,Y,Z)]RPS = cycle dependent maximum allowable design peaking q

factor which ensures that the centerli fuel melt limit will be preserved for all operation. (

(X,Y,Z)]RPS includes allowances for calculational a d measurement uncertainties.

I((X,Y,Z) = the design power distributions for Fn. (le (X,Y,Z)is provided in Table 2 for normal operation and tab escalation testing during initial startup.

$1 (X,Y,Z) = the margin remaining to the CFM lindt in core location C

X,Y,Z from the transient power distribution. Mc(X,Y,Z) calculations parallel the Mg(X,Y,Z) calculations described in DPC NE 201 IPA, except that the LOCA limit is replaced with the CFM limit. M (X,Y,Z) is provided in Table 4 for C

normal operation and table 4A for power escalation testing during initial startup.

UMT = Measurement Uncertainty (UMT = 1.05).

MT = Engineering Hot Channel Factor (MT = 1.03).

TILT = Peaking penalty that accounts for allowable quadrant power tilt ratio of 1.02.

b

[I (X,Y,Z)]RPS is similar to the parameter identified as N (X,Y,Z) in NOTE:

g g

DPC NE 2011PA except that M (X,Y,Z) replaces Mg(X,Y,Z).

C

,dI,6 KSLOPE = 0.078

)Jk where KSLOPE = Adjustment to the K g value from OTAT required to compensate for eaf 19e that (E(X,Y,Z)]RPS exceeds it limit.

n s

2 60

+

um.

i C-6/

i I

CNEl.0400 74 (kev. 0) j 16of302 1

l Cataw ba 1 Cycle 7 Core Operating 1.imits Report

.26 Nuclear Fnthniny Rise 1101 Channel Factor. F31g(X.Yli (Einecif1 cation l

J.7 3a n I

)

i The following parameters are required for core monitoring per the LCO Requirements of j

Specification 3/4.2.3:

i j

2T.1 (Fag (X,Y))LCO = MARP (X,Y)

1.0 +

(1.0 - P) i

.37 where (MARP(X,Y)) = Catawba 1 Cycle 7 Operating Limit Maximum Allowable j

Radial Peaks. (MARP(X,Y)) is provided in Table 1.

}

' Thermal Power j

P = ' Rated Thermal Power i

i The following parameters are required for core monitoring per the Surveillance l

Requirements of Specification 3/4.2.3:

F$g(X,Y)

May(X,Y) 56.2 (fig (X,Y)]SURV.

i 3?

where (Ikg(X,Y)]SURv = cycle dependent maximum allowable design peaking factor which ensures that the Fan (X,Y) limit will be pjr served for operation within the LCO limits.

l

[F3n(X,Y)]SURV includes allowances for calculational i

and measurement uncertainties.

i F$g(X,Y) = the design power distribution for Fag. F$g(X,Y) is provided in Table 5 for normal operation and table SA for power escalation j

testing during initial startup.

M g(X,Y) = the margin remaining in core location X,Y to the Operational 3

DNB limit in th: transient power distribution. MAH(X,Y)is

- provided in Table 6 for normal operation and table 6A for power escalation testing during initial startup.

j UMR = Uncertainty value for measured radial peaks, (UMR = 1.04).

TILT = Peaking penalty that accounts for allowable quadrant power tilt l

l ratio of 1.02.

l 2-61' l'

l-i

... _ = -..

1 C-61 I

CNEl 0100 24 (Rev. 0) 17 0f 302 Catawba 1 Cycle 7 Core Operating Limits Report i

1 NOTE: ((g;(X,Y)]SURV is the parameter identified as [gf(X,Y) in DPC NE-i 2011PA.

j A6.3 RRil = 3.34

.sn j

where RRif = Thermal Power reduction required to compensate for each 1% that Fy;(X,Y) exceeds its limit.

j 4

l

,2.4.4 TRII e 0.04 l

39 where TRH -Reduction in OTAT K setpoint required to compensate for each 1%

j-i

)

that Fat (X,Y) exceeds its limit.

l l

Inse,+ 4 -+

i l

1 i

e I

4 2-62

l C 43 i

l i

j Insert 4 -

3.N lloron I)llution Milleation Sntem (Snecifica lon 3/4.3.3.12) e 3.8.1 Reactor Water Makeup Pump flowrate limits:

}

Anolicable Mode Limil Mode 3 or 4 s 150 gpm a

Mode 5 s 70 gpm i

1 i

3.9 Accumulators (Sueciflention 3/4.5.1) l-3.9.1 11oron concentration limits during modes: 1,2 and 3:

Parameter Lhnits i

Cold Leg Accumulator minimum bomn 1,900 ppm l

concentration for LCO 3.5.lc 1

i Cold Leg Accumulator maximum baron 2,100 ppm i

concentration for LCO 3.5.lc i

Minimum Cold Leg Accumulator boron 1,800 ppm concentration required to ensure post LOCA l

suberiticality 1

d i

3.10 Refueline Water Stornee Tank (Snecifiention 3/4.5.4) 3.10.1 Boron concentration limits during modes: 1,2,3 and 4:

i Parameter Limits 4

Refueling Water Storage Tank minimum boron '

2,000 ppm i

concentration for LCO 3.5.4b i

Refueling Water Storage Tank maximum boron 2,100 ppm-concentration for LCO 3.5.4b 1

i l-2-63 i-

.-.-_,-__._.--_-.-_-.m.._-_-...

C-6,4 j

litsert 4 - colitiittled 3.11 Instrumentation (Soccif1 cation 3/4.9.21 1

l 3.11.1 Reactor Makeup Water Pump Flowrate Limit:

a l

Aonlicable hhxle Lhnit Mode 6 570 ppm i

i A

I i

1 4

4 I

i f

i 2-M

C-4 s i

Technical Justification (Catawba) 4

'l l

1 i

h i

i i

1

-I t

a

. ~ _. _. _ _

C-H inged Pelocation of Certain itemn f rom t he Tecimical Cog ificatient, to the COLR Sevetal of the f ollowing proposed Technical Specification revisiono at e i

relocationn of cer tain items f rom t he Technical Speci fications to the Core Operating Limitn Report (COLR).

The justification for these changes is, to a large extent, common to all of them.

The following paragraphs describe this common justification to avoid reretition.

Technical Justificatien for Proooced Relocationa from Technical 1

Soecifications to COLR Catawba Facility Operating hicense amendment numbers 74 and 68 dated

!!ay 17, 1990 for units 1 and 2 respectively, revised the Catawba Technical Specifications to replace the values of certain cycle-specific pat amet er limits with a ref erence to the COLR, which contains the values of the limits, However, additional existing cycle-specific parameter limito in the Catawba Technical Specifications, not included in the above amendments, will have to be revised due to changes in these parameters in support of Unit 2 Cycle 6 operation and reload j

i design.

Similar limito have also changed in recent McGuire fuel cycles.

In addition, McGuire Facility Operating License amendment numbers 105 and 87 dated March 15, 1990 for units 1 and 2 respectively, revised the McGuire Technical Specifications to incorporate an identical COLR methodology, Therefore, in order to simplify tJRC review i

of identical Technical Specification revision proposals, it is proposed that the McGuire Technical Specifications be changed identically with respect to the applicable item relocations to the COLR, In t ecognition of the burden on licensee and tJRC resources associated with changes to Technical Specifications, the 11RC isoued Generic Letter 2

88-16 on October 4, 1988 encouraging licenoees to propose changes to Technical Specifications that are consistent with the guidance provided in the enclosure to the generic letter, This enclosure provides guidance for the preparation of a license amendment lequest to modify Technical Specifications that have cycle-specific parameter limits.

With the implementation of this alternative the tJRC concluded that reload license amendments for the sole purpose at updating cyc10 specific parameter limits would be unnecessary.

The proposed revisions described below would relocate the cycle-specific parameter limits from j

the Catawba Tachnical Specifications in accordance with the guidance provided in the enclosure to Generic Letter 88-16.

Prococed Revision to Technical Soeci fication Table 2.2-1 It is propoced that the following Technical Specification setpoints be relocated to the Core Operating Limits Report:

overtemperature AT reactor trip setpoint, K1 e

Overtemperature AT reactor trip heatup setpoint penalty coefficient, K2 Overtemperature AT reactor trip depressurication setpoint e

penalty coefficient, K3 Overpower AT reactor trip setpoint, K4 e

Overpower AT reactor trip heatup setpoint penalty coefficient, e

l K6 l

Measured reactor vessel AT lead / lag time constants 11 and 12 l

2 l

s

C-07 i

e Measured AT lag time constant 13 e Measured teactor vessel average temperatute lead / lag time

]

constants 14 and 15 e Measured t eact or vessel average temperature lag time constant 16 e Measured 2 eactor vessel average temperature rate / lag time i

j constant 17 e

f 1(AI)

positive" breakpoint, i.e.,

the moct positive imbalance (axial flux difference) at which no overtemperature AT reactor trip setpoint penalty is required due to chewed axial power chapes e

f1(AI)

negative' breakpoint, i.e.,

the most negative imbalance at which no overtemperature AT reactor trip setpoint penalty is required due to skewed axial power shapes e

f1(AI)

positive" slope, i.e.,

the rate at which an I

overtemperature AT reactor trip setpoint penalty is applied for axial power shapes skewed to the top of the core e

f1(AI)

negative

slope, i.e.,

the rate at which an i

overtemperature AT reactor trip setpoint penalty is applied for skewed axial power shapes skewed to the bottom of the core e

f2(AI) " positive" breakpoint, i.e.,

the most positive imbalance (oxial flux difference) at which no overpower AT reactor trip setpoint penalty is required due to skewed axial 1

power chapes e

f2(AI)

negative" breakpoint, i.e.,

the most negative imbalance at which no overpower AT reactor trip setpoint penalty is required due to skewed axial power shapes e

f2(AI)

positive' slope, i.e.,

the rate at which an overpower AT reactor trip setpoint penalty is applied for axial power shapes skewed t o the top of the core e

12(AI) " negative" slope, i.e.,

the rate at which an overpower AT reactor trip setpoint penalty is applied for skewed axial power shapes skewed to the bottom of the core Technical Just ificatim The setpoints listed above characterize the Reactor Protection System trip functions which protect the reactor core from departure from nucleate boiling (DNB) and centerline fuel melt (CFM).

The f(AI) setpoints are calculated by the methodology described in Chapter 4 of Duke Power Company topical report DPC-NE-20ll-P-A, ' Nuclear Design Methodology for Operating Limits of Westinghouse Reactors'(Reference 2),

This report was approved on January 24, 1990 and is listed in Section 6.9.1.9 of t he Catawba Technical Specifications.

Chapter 5 of Duke Power Company topical report DPC-NE-2004, "McGuire and Catawba Nuclear Stations Core Thermal-Hydraulic Methodology Using VIPRE-Ol',

discusses the relationship of the AT reactor trip setpoints to the core safety limits.

This report was approved on November 15, 1991.

As described in Chapter 4 of DPC-NE-2011-P-A, thermal margin calculations are performed for each fuel cycle to verify that the axial flux difference reactor trip penalty function envelopes, f 1(AI) and i

i 3

l l

O*N fptAI), remain conservative.

If the proposed Technical Specification revision in approved, inst anwu in which thit check fails, but fct which diffotent envelopes can be demonstrated to both 1) provide adequat e pret ect ion against licensing basio accident s and 2) be cat is f led by t he fuel cycle design under considetation, no Technical Specification change would be necessary.

An described in Chapter 2 of Duke power Company topical report DPC-NE-9te ;uire/ Catawba Nuclear St ation !!ultidimensional Reactor 3001-PA, r

Transients and Saf ety Analysis Physics Parameters Methodolog)"

I (referenco 4), safety analysis physics parameters are checked each cycle to determine that the assumptions of the FSAR Chapter 15 accident analyses remain valid.

If the proposed Technical Specification revision is approved, instances in which this check fails for a particular parameter, but for which a difielent parameter value can be demonsti at ed to bot h 1) yield an acceptable result when the applicable licensing basis accident (s) are reanalyzed and 2) be satisfied by the fuel cycle design under consideration, no Technical Specification change would be necessary.

Reanalysis would be likely, in order to get an accept able answer, to use revised

'K' and

't' values.

Reanalysic in this case would also, of course, be performed according to the methodologies approved in DPC-NE-3001-PA and/or Duke Power Company topical report DPC-NE-3002-A, 'McGuire/ Catawba lluclear Station FSAR Chapter 15 System Transient Analysis Methodology".

DPC-NE-3001-PA and DPC-NE-3002-A were approved November 15, 1991 and are listed in Section 6.9.1.9 of the Catawba Technical Specifications.

i The pr oposed Technical Specification change is similar to those approved on Sept ember 16, 1991 as Facility Operating License amendments J

191, 191, and 188 for Oconee Units 1, 2, and 3, respectively.

In these amendment s t he Reactor Protection System setpoints f or the Flux / Flow /Imb31ance and Variable Low Reactor Coolant System Pressure trip functions were relocated to the COLR.

As stated in the Oconee Technical Specifications, these trip f unctions provide equivalent p2 0tection against f uel thermal limit s, specifically DNB and CFM, as the catawba overtemperature AT and overpower AT reactor trip functions.

Recent instances where one or more of these setpoints has been changed in a Teclulical Specification revision proposal include McGuire 1 Cycle 8.

McGuire 2 Cycle 8, Catawba 1 Cycle 7, and Catawba 2 Cycle 5.

ProDM "d Revision tr_s Minimum Poron Concent ration Limit s on ECCS Water Egurces It is proposed that the following minimum boron concentraticn limits be relocated from the Technical Specifications to the COLR:

The cold leg accumulator minimum boron concentration limit in Technical Specification 3.5.1.c The Refueling Water Storage Tank (RWST) minimum boron e

concentration limit in Technical Specification 3.5.4.b Technical Justification The limits on minimum boron concentration for the accumulators and RWST are verified each cycle to ensure the safety analysis assumptions for these parameters remain valid.

The minimum boron concentration limits ensure the reactor will remain subcritical during a Loss of Coolant 4

6

C_- G 9 J

1 l

Accident (Reference 11, Chapter 15.6.5.2 ' Post LOCA Subcriticality Evaluation *).

The post LOCA subcriticality evaluation ptoviden an available sump mixed mean boron concentration curve that must bound the required cold, post LOCA critical boron concentrationc for each cycle.

The critical boron concentration requirements for each cycle are determined using the methodology described in Section 9 of DPC-NF-2010A (Reference 14).

In addition, the minimum boron concentration will be evaluated each cycle to ensure the solution recirculat ed within containment after a LOCA will be maintained at a pH value of G 9.5.

This pH limit minimizes hydrogen production from the corrosion of j

aluminum in containment and reduces the effect of caustic stress corrosion on mechnical systems and components (Reference 11, Section 6.1.1.2.2).

The pH band given in the bases for Specification 3/4.5.4 in the Catawba Technical Specifications and Specification 3/4.5.5 in the McGuire Technical Specifications is revised to provide consictency between the values at both plants while meeting the NRC criteria for sump pH af t.er a LOCA, contained in Branch Technical Position MTEB 6-1.

The pH band is revised to be 2 7.5 and 5 9.5 for both plants.

The ice condenser ice bed pH ia maintained between 9.0 and 9.5 (Specification 3.6.5.1(a)) and is the most alkalina contributor to the containment sump pH after a LOCA, therefute, the pH will be limited ty the maximum pH limit of the ice bed.

The lower limit on the band is chosen such that stress carrosion cracking will not occur for an extended period I

following a LOCA and iodine retention in the containment sump water is enhanced.

A similar Technical Specification revision was approved j

January 5, 1993 for Oconee Units 1, 2, and 3, Amendment Nos., 197, 197, and 194, respectively, to move these cycle-specific values to the COLR.

The boron concentration value for the Oconee Core Flood Tank (CFT),

which performs the same function as the Catawba Cold Leg Accumlators (CLA), was proposed tc be relocated to the COLR.

Also, the, boron concentration value for the Oconee Borated Water Storage Tank (DWST),

which performs the same function as the Catawba Refueling Water Storage i

Tank (kWST), was proposed to be relocated to the COLR.

Preposed Revision to Maximum Doron Concentration Limits on ECCS Water Sources It is proposed that the following maximum boron concentration limits be relocated from the Technical Specifications to the COLR:

The cold leg accumulator maximum boron concentration limit in Technical specification 3.5.1.c The Refueling Water Storage Tank maximum boron concentration i

e 1

limits in Technical Specification 3.5.4.b Technical Jurtification The limits on maximum boron concentration for the accumulators and RWST 1

ar e based on the minimum boron concentration limits, which are determined using the methodology in DPC-NF-2010A (Reference 14), to provide adequate plant operating space between these limits.

They are also evaluated to ensure boron precipitation is precloded followin'ga LOCA (Reference 11. Section 6.3.3).

The boron precipitation analysis uses methods and assumptions described in Westinghouse letter CLC-NS-309 dated April 1, 1975 with the principal input parameters given in Table 6-99 of Reference 11.

In addition, the maximum boron concentration will be evaluated each cycle to ensure the solution 5

i

CAO recir culat ed wit hin cont ainment after a LOCA will be maintained at a pil value ? 7.5.

This pil limit minimiter the evolut ion of iodine and reduces the effect of chloride stress corrosion on mechnical cystems and componentc.

Iteneced Revicicn to Minimum Volume & Doron Concentratien Limite on Porat ed Water Sourceo f or Reacter Shutdown It is proposed t hat the following minimum volume and boron concent ration limit s be t elccated f rom the Technical Specifications to t he COLR:

The 13oric Acid Storage System minimum volume and boron w

concentration limits in Technical Specification 3.1.2.5 &

3.1.2.6 e The Refueling Water Storage Tank minimum volume and boron concentration limits in Technical Specification 3.1.2.5 &

3.1.2.6 Technica1 Juctificatinn The limit s on minimum volume and boron concentration f or the Boric Acid Storage System and RWST are verified each cycle to ensure the core chutdown analysis assumptions for these parameters remain valid.

The minimum volume and boron concentration limits ensure that negative reactivity control is available during each mode of plant operation.

The minimum volume and baron concentration requirements are based on those that will provide a 1% Ak/k shutdown margin for temperatures less than or equal to 200 oF and 1.3% Ak/k for temperatures greater than 200 F at all times during a given cycle.

The required boron concentratiens are determined using t he methodology described in Section 4 of DPC-NF-2010A (Reference 14).

Further, the limits on RWST minimum volume and boron concentration help ensure the reactor will temain subcritical during a Loss of Coolant Accident (Reference 11.

Chapter 15.6.5.2 ' Post LOCA Subcriticality Evaluation").

In addition, the limits on RWST minimum volume and boron concentration are evaluated each cycle to ensure the solution recirculated within containment after a LOCA will be maintained at a pli value s 9.5.

This pli limit minimizec hydrogen production f rom the corrosion of aluminum in containment and reduces the offeet of caustic stress corrosion on mechnical systems and components.

The pH band given in the bases for Specification 3/4.1.2 in the Catawba and McGuire Technical Specifications is revised to provide consistency between the values at both plants while meeting the NRC criteria for sump pli af ter a LOCA, contained in Branch Technical Position MTEB 6-1.

The pil band is revised to be 2 7.5 and s 9.5 f or both plants.

The ice condenser ice bed pH is maintained between 9.0 and 9.5 (Specification 3.6.5.1(a))

and is the most alkaline contributor to the containment sump pH after a LOCA.

Theretore, the pH will be limited by the maximum pH limit of the ice bed.

The lower limit on the band is chonen such that stress corrosion cracking will not occur for an extended period following a LOCA and iodine retention in the containment cump water is enhanced.

The value fcr the RWST minimum contained borated water volume prescribed in Specification 3 /4.1.2. 6 (b) (1) is baced on the ECCS RWST volume requirements in Specification 3/4.5.4 and is a larger volume 6

b

=.

CL '7 /

i j

than required for reactivitity control, as shown in the bases for j

Specification 3/4.1.2.

The minimum RWST volume report ed in t ha COLR will be the volume required to maintain reactivity control while i

Speci f ication 3 / 4.1. 2. 6 (b) (1) is changed to refer to the value in Specification 3/4.5.4(a) or the Cohk, whichever is larger.

In addition, the bases f or the ECCS RWST mimimum cont ained borated water volume, currently located in the reactivity control bases, is relocated j

to the ECCS bases.

The Catawba Teclutical Specification baces f or the allowances assumed in the Boric Acid Storage Tank and RWST minimum

~

contained water volumes are revised for clarification.

A similar Technical Speci fication revision was approved January 5, 1993 for Oconee Units 1, 2, and 3, Amendment IJos., 197, 197, and 194, i

respectively, to move these cycle-specific values to the COLR.

The volume and baron concentration value for the Oconee Concentrated Doric Acid Storage Tank (CBAST), which performs the same function as the Catawba Boric Acid Tank, was relocated to the COLR.

Also, the boron concentration value for the Oconee Borated Water Storage Tank (BWST),

which performs the same fuaction as the Catawba Refueling Water Storage Tank (RWST), was relocated to the COLR.

i i

Procesed Revicion to Reactor Makeue Water Pumn Plewrate Limits It is proposed that the following reactor makeup water pump flowrate limits be relocated from the Technical Specifications to the COLk:

The reactor makeup water pump flowrate limits in Technical Specification 3.3.3.12(a)(2),

3. 3. 3.12 (b) ( 2 ) & 4.3.3.12.2(b)

The reactor makeup water pump flowrate limits in Technical 3

e Specification 3.9.2.l(a)(2) & 4.9.2.1.2(d)

Technical Juctification Catawba is equipped with a Boron Dilution Mitigation System which serves to detect uncontrolled dilution events in Modes 3 - 6 of plant

]

operation.

The BDMS uses two source ange det ectors to monitor the subcritical multiplication of the reactor core.

An alarm setpoint is continually calculated as four times the lowest count rate, including compensation for background and the statistical variation in the count rate.

Once the alarm setpoint is exceeded, each train of the BDMS will automatically shut off both reactor makeup water pumps, align the suction of the charging pumps to highly borated water from the Refueling Water Storage Tank, and isolate flow to the charging pumps from the Volume Control Tank, since these functions are automatically actuated by the BDMS, no operator action is necessary to terminate the dilution event and recover the shutdown margin.

In the event one or more trains of the BDMS is inoperable, the reactor makeup water pump flowrate limits ensure that the operator has sufficient time to recognice and terminate a boron dilution event prior to the loss of shutdown margin during each appropriate mode of plant operation, Each cycle, a bounding ratio of initial to critical boron concentration is established from the reload design.

This ratio is used to calculate the maximum reactor makeup water pump flowrate which satisfies the operator action time acceptance criteria of the Standard Review Plan.

The limits on reactor makeup water pump flowrates when the Baron Dilution Mitigation System (BDMS) is inoperable are verified each cycle to ensure the safety analysis assumptions for these parameters remain valid.

When the calculated reactor makeup water flowrate is found to be less than the existing flowrate limits, the flowrate limits must be reduced such that the operator action time acceptance criteria can be 7

m__

]

C.'72 i

j me t..

These cycle-specific parameter litnits are verified using the Imc q

appt oved tuethodology pr ovided in t ho at t achment t o a Duke Power le't er j

to the U.

S. !Juclear Regulat.ory Cownission, Supplementary

{

Information Relative to Topical Report BAW-10173; Boron Dilution Analysia", dated May-15, 1991 (Reference 12) and Catawba FSAR l

(Reference 11) Section 15.4.6.

Therefore, the cycle-specific limits l

have been relocated to the COLR.

i 1

1 i

1 2

1 I

i s

4 1

i i

4 i

I a

i 2

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1, 1

i i

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3 1

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8 4

x--~.n,-,w.-,2-,.-,

-v---ww

.n,,-.,.

r

,,,-r,

,.-,.a

~-.-,,-em-ow,.w w wm

-m-

.me,-.w

--w,ve..m.,ww

,,e.,-,,,n.,,,-,,,,,,-,,ww.vn-,,-,.--~~myw-,-,,w,,.-~,,,--,ee-w,

. e v e--

com 1

i 1

l Generic Letter 88-16 provides guidance fur the removal of cycle-j specific parameter limits from the Technical Specifications.

Amendment s 74 and 68 dated May 17, 1990 for units 1 and 2 1

respectively, revised the Catawba Technical Specifications to l

replace the values of certain cycle-specific parameter limits with j

a reference to the Core Operating Limits Report (COLR), which contains the values of the limits.

Since approval of this i

amendment, additional cycle specific parameters have been j

identified due to revisions which were necessary to support Catawba Unit 2 Cycle 6 operation and past McGuire reloads.

l NO SIGNIFICANT HAZARDS EVALUATION i'

10 CFR 50.92 states that a proposed amendment involves no significant hazards consideration if operation in accordance with the amendment would nots i

l 1)

Involve a

significant increase in the probability or consequences of an accident previously evaluated; or i

2)

Create the possibility of a new or different kind of accident from any accident previously evaluated; or j

3)

Involve a significant reduction in a margin of safety.

This proposed amendment will not increase the probability or consequences of an accident which has been previously evaluated.

The cycle specific parameters which have been identified for relocation to the COLR will be calculated using NRC approved methodology and the Technical Specifications will continue to j

require operation within the cycle specific parameters.-

For the above reasons this amendment is considered administrative, and does not increase the probability or consequences of an accident previously evaluated.

Operation in accordance with this proposed - amendment will not j

create any failure modes not bounded by previously evaluated i

accidents.

Therefore, this change will not create the possibility i

of a new or different kind of accident from any kind of accident previously evaluated.

t l

The Technical Specifications will continue to require operation within the bounds of the cycle-specific parameter limits.

The cycle-specific parameter limits will be calculated using NRC 1

approved methodology.

In addition, each future reload will require a 10 CFR 50.59 safety-review to assure that_ operation of-the unit i

within the cycle-specific limits will not involve a reduction in a margin of safety.

Therefore, no margins of safety are affected by i

the relocation of cycle-specific parameter limits _to the COLR.

l Besed on the

above, Duke has concluded that there are no significant hazards considerations involved in this amendment m

,.--m.-

we - - -, -

{

O-7Y i

j request.

l The proposed Technical Specification change has been reviewed 1

against the criteria of 10 CFR 51.22 for environmental j

considerations._

As shown above, the proposed change does not involve any significant hazards consideration, nor increase the j

types and amounts of effluents that may be released offsite, nor increase the individual or cumulative occupational radiation exposures.

Based on this, the proposed Technical Specification j

change meets the criteria given in 10 CFR

51. 22 (c) (9 )

for categorical exclusion from the requirement for an Environmental Impact Statement, a

a J

a l

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h

.,,.-,..,---,.~..,w y...,..._4

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f C.- 15 I

s I

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1 2

1 1

i 7

-l.

i I

i 3

i i

k.

1 i

Changes to the Catawba FSAR 4

i I

l 9

1 4

6 i

i t

f e

I r

I 2-110

c.-%

Catawba Nuclear Station 7.2 lleactor Irip System one of the two intermediate range channels reads above the P.6 setpoint and is automaticauy reinstated when both intermediate range channels decrease below the P 6 setpoint value. This tnp is also autornatically bypassed by two out of four logic from the power range protection interlock (P 10).

This trip function can also be reinstated below P 10 by an administrative action requiring rnanual actuation of two control board mounted switches. Each switch will reinstate the inp function in one of the two protection logic trains, The source range trip point is set between the P 6 setpoint (source range cutoff power level) and the maximum Source range power level.

The channels can be individually bypassed at the nuclear instrumentation racks to permit channel testing during plant shutdown or prior to startup. This bypass action is annunciated on the control board.

d. Power range high positive neutron flux rate trip This circuit trips the reactor when a sudden abnormal increase in nuclear power occurs in two out of four power range channels. This tnp is available for rod ejection accidents of low worth from mid power and is always active.

e. Power range high negative neutron Dux rate trip This circuit trips the reactor when a sudden abnonnal decrease in nuclear power occurs in two out of four power range channels. This trip provides protection against two or more dropped rods and is always active. Protection against one cropped rod is not required to prevent occurrence of DNDR per Section 15.2.2:" Loss of Extemal Load" on page 15 33.

Figure 7 2. Sheet 3. shows the logic for all of the nuclear overpower and rate trips.

2. Core Thermal Overpower Trips The specific inp functions generated are as follows:

a. Overtemperature AT trip This trip protects the core against low DNDR and trips the reactor on coincidence as listed in Table 71 with one set of temperature measurements per loop. The setpoint for this trip is continuously calculated by analog circuitry for each loop by solving the following equation (for which the dnail+44me comtriere giwaNn Table 2.21 of the Technical Specificationst Pntame1w nlues nec prescribed

{,

T.,,

- T *,,, + K (P - 2235)-M 6Twipng-aTo K-K y

i 3

is)

Indicated aT at rated thermal power ATo

=

Average reactor coolant temperatrue ('F)

T.,,

=

T',,,

Indicated T.,, at rated thermal power

=

Pressurizer pressure (psig)

P

=

Preset bias K

=

i Preset gain which compensates for effects of temperature on the DNB K

=

limits Preset gain which compensates for the effect of pressure on the DND K

=

3 limits Preset time constants t h,t6

=

i Laplace transfctrn operator (seconds 1) 5

=

A function of the neutron Oux difference between upper and lower long R3%

=

[f(LM ion chambers. (Refer to Figurc 7 3)

A separate long ion chamber unit supplies the flux signal for each overtemperature aT trip channel.

2-111 (01 OCT 1991) 7 21

C 't1 7.2 Reactor Trip S sicm Catawba Nucicar Station 3

9 lacteases in as beyond a pre-defined deadband result in a decrease in trip setpoint. Refer to Figure 71 He required one pressunzer pressure parameter per loop is obtained from separate sensors connected to three pressure taps at the top of the pressurizer. Four pressurizer pressure signals ne obtained from the three taps by connecting one of the tapt to two pressure transmitters.

Refer to Section 7.2.2.13 " Pressurizer Pressure" on page 7 40 for an analysis of this arrangement.

Figure 7 2 Sheet 5, shows the logic for overtemperature T trip function.

b. Overpower AT trip his inp protects against excessive power (fuel rod rating protection) and trips the reactor on coincidence as listed in Table 71, with one set of temperature measurements per loop. The setpoint for each channel is continuously calculated using the following equation (for which the krikd timsonstants-are-pha in Table 2.21 of the Technical S ecifcations):

r-7%rnme.w-vaine.s pre preStrbed ar

= ar. }x. - xt( i :',,, )( J,, )(r..s)-

ys( +c )T..s - T..s

. -m

&(av) where:

AT, Indicated AT at rated thermal power

=

Tt64 A function of the neutron flux difTerence between upper and lower long

=

{(Ap) >

ion chamber section.

K 4 A preset bias

=

Ks A constant based on the effect of rate of change of T.,, on overpower AT limits K.

A constant based on the effect of T.,, on overpower AT limits

=

T',,,

Indicated T.,, at rated thermal power

=

T.,,

Average reactor coolant temperature ('F)

=

Preset time constant (seconds)

v.,t,

=

laplace transform operator (seconds 1) s

=

The source of temperature and flux information is identical to that of the overtemperature AT trip and the resultant AT setpoint is compared to the same AT Figure 7 2, Sheet 5, shows the logic for this trip function.

3. Reactor Coolant System Pressunzer Pressure and Water Level Trips The specific tnp functions generated are as follows:

a. Pressurizer low pressure trip The purpose of this trip is to protect against low pressure which could lead to DNIl. The parameter being sensed is reactor coolant pressure as measured in the pressurizer. Above P 7 the reactor is tnpped when two out of four pressurizer pressure measurements (compensated for rate of change) fall below preset limits. This inp is blocked below P 7 to pennit startup. The trip

' logic and interlocks are given in Table 71.

He trip lope is shown on Figure 7 2, Sheet 6.

b. Pressurizer high pressure trip he purpose of this trip is to protect the Reactor Coolant System against system overpressure.

He same seasors and transmitters used for the pressurizer low pressure trip are used for the high pressure trip except that separate bistables are used for inp.

These bistables trip when 2-112 7 22 (01 OCT 1991) l

_ _... _ _ - _ _. _ _ __. _ _ _ _ _. _ _ _ _ ______.. _ _ =.. _. _ _ _ _ _ _ _ _ _.. _. _ _.

4 4

4 h

4 1

J i

i-4 o

l 1

4 1

1 a

4 1

i e

J l

1 e

l e

j e

k 5

3 i

1 i

e i

t 5

d i

P I

l l

l I

i I

i l

l l

hi-l l

Changes to Technical Specifications (McGuire) 9 2 65

t r

TABLE 2.2-1 (Continued) l E

5.

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 2

NOTATION E

d NOTE 1: DVERTEMPERATURE AT w

E (aT/AT,) (f

) (1 + r35) I El 2 I+r53 0(1 + r S)-T'] + K (P-P') - f (al)

I

~E 3

y s

to Where:

AT Heasured AT by Loop Narrow Range RID

=

AT Indicated AT at RATED TilERMAL POWER,

=

g f

Lead-lag compensator on measured AT,

=

'?

11. I2

= Time constants utilized in the lead-lag controller for AT, :12 8 sec, :2 S ? sec.

as presented in the Ca e Opraf;3 ;,.h Repd; L

c 1

Lag compensator on measured AT,

=

y, 7, 13 Time constants utilized in the lag compensator for AT,13

? sec.* ats prese,sted in

=

the Core. O a<nt:,, L;r,,:Is Repet; p

y K

1.1958,=Overtempe f~ce bT reue% +r p seF ;d ss prese<ied h f~e Core Opack 3 Cdh R-fxt, 3

p tg K

over%perstare. sf reae% f<;p kes% setp:nt peaalFy coeff;elmt as pres enfeli.,

0.03143

=

f g

l E, the Core Opera %3 Ll~,h Report; yy 17s3 The function generated by the lead-lag controller for T,yg dynamic compensation,

=

Time constants utilized in the lead-lag controller for I,yb, L; t4 13

=

4 2 28 ccc, is ' 1 Sec., as fctseded ln % Core O ra 1

p 3

,.,, b R e p e t, y

33;;

T Average temperature, F,

=

"~

1

~

Lag compensator on measured T

=

1+rSs avg' N

t

IABIE 2.2-1 (Continued)

}

z

'l c5 REACIOR TRIP SYSTEM IllSTRUMEt1TATI0rl TRIP SETP0lt4TS fiOTATI0tl (Continued)

C5 l1010 1:

(Continued) 7.

Time constant utilized in the measured I lag compensator, q ' ? sec as preseided m

=

g is 1he Core Operd:~3 L;,.ts Rcrect, o,

3 T'

5 588.2 F Reference I at RATED TilERMAL POWER,

=

a avg K

" 0 00

O * "P * #"

"" "' E f # * * #'

C#*

3 geseded in +Ac. Core. Opern+;n3 L;mlb R, pet, P

f ressurizer pressure, psig.

=

P' 2235 psig (tiominal RCS operating pressure),

=

Laplace transform operator, sec 8,

S

=

rf and f (AI) is a function of the indicated difference between top and bottom detectors so y

of the power-range nuclear ion chambers; with gains to be selected based on measured S

instrument response during plant startup tests such that:

between[" -' " ^** '- f (al) = 0, where q-r4e?pos/R,< ~.J NeyJirc ' f,(AI) bredpints f

3 (i) for q

~9 and q are percent RATED Regr; t

b 3

t b

TilERMAL POWER in the top and bottom halves of the core respectively, and q

+q t

b gg is total TilERHAL POWER in percent of RATED Ti!ERMAL POWER; y*

the. f,(AI)%es%* breek@d prese.tel en he Core Operatl= Limah Rep -T -

b gg (ii) for each peYcent imbalance that the magnitude of q -

s m re negative than 3Ta!,

b am g 3, the AT Trip Setpoint shall be automatically reduced by 5.153% of aTo, and 4ke f(hi)"nega+we" slope prm2el,', & C.<e Opa& GL z2 2,o (iii) for each percent imbalance that the magnitbde 6f q -

is more positive than 0.0%al,theATTripSetpointshallbeautomatica1kyreucedby,1.511TcfaTo bthe f,(at)*pashe bredp:J preseded o'n He (Se. f,(LE) "peshe!' slope presched in +ke G.s C re ope rd:.,3 L;, :h Rwat Core oger.b3 L;, h Repet MH ev t

N

1 b'

L i

i i

x TABLE 2.2-1 (Continued) j

,i

.y REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SE1 POINTS NOTATION (Continued) i

-c-i5 NOTE 2:

OVERPOWER AT.

-4 1

I (AT/AT,) (3

) (1 + T 5) $ K4 5

l+

2

-K I

5) Il + r S) 6 UIl + r S)- T1 - f (0 )

s 2

3 s

LE t

to Where:

AT As defined in Note 1,

=

1 AT, As defined in Note 1, l

}

=

I*

, y,7]

= As defined in Note I ti, T2

= As defined in Note 1 m-1+T5

= As defined in Note 1 I

3 i

'I K

5 1.0300, = Overgaaee bi~reador te,y.sefp,~ t as presented ls fle Can O erdn3 L;n,h Reper, 4

p S

0.0W for jacreasig average temperature and 0 for decreasing average l

K

=

temperature, g

15 7

= The function generated by the rate-lag controller for T,yg dynamic 1+157

'}g compensation,

'ES Time constant utilized in the rate-lag controller for T 'O, !; ~' S :^c, as ore 3eated,%

=

17,

' g gf

+,e core. Operak Lion,h Repet; 3

As defined in No 1,

=

1 + Ts5 w-As defined in Note 1,

-ts

=

m

?:P

7 K

= 0.001230/ F for T > T" and K6 = 0 for T 1 I",

6 3g Overgoaec AT renehr it,'y hestap serp;nt genalty coefflele,8 as prese ded

,, the Care opc&3 L;,,,:ts Rep,rr y

b

j i

I y

TABLE 2.2-1 (Continued) l aS' REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS M

I NOTATION (Continued)

[

e E

I i

d T

As defined in Note 1, L

=

w l

a, I"

2

-< 588.2 F Reference T at RATED THERMAL POWER,

=

avg a

i l

S As defined in Note I, and l

=

m i

l j

f (AI) is a function of the indicated difference between top and bottom detectors

[

2 of the power-range nuclear ion chambers; with gains to be selected based on n:easured instrument response during plant startup tests such that:

t between 25 }. and :25% a.; f (AI) = 0, where qs.&c*aa!'nep%* ft(bT) bredp;2s as p) g +4e (i) for q q

2 and q are percent RATEDTHERMLPOWERinthetopandbottomhalvesofthecorerkspectihely,and l

m q

9 is total THERMAL POWER in percent of RATED THERMAL POWER; t

b

-G(br) *neytlse* brec.L e:nt presented on ff e Core Operat:3 Cn.,fr Report

?

de p

e, (ii) for each percent imbalance that the magnitude of q q is more negative than

}

--3Sfr-ah the AT Trip Setpoint shall be automatically redbced by 7.L" of aTo; and

[

y 6 (41) % d:ve." skye. presa,ted ;n -the Core 'opr.!:3 Cm:ts Report 2 the 3

j i

(iii) for each percent imbalance that the magnitude of q q is more positive than l

35%AI,theATTripSetpointshallbeautomaticallyredbcedby7.0%ofATo.

l 0 de f(ar) *pos,be" break o:a preseJed ln he Core operJ The channel's maximum Trip Setpoint shall not exceed its comp % C.,:fs Rep.<t t

p uted Trip Setpoint by more than f

3 3,,-

Note 3:

5 88 3.6%.of Rated Thermal Power.

[

i EE R2 Note 4:

The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 4.2%

f l

5g of Rated Thermal Power.

l i

zz the. ((bi)"pos%"3 ope preseded on tke. Oore. O ed:~3 C m.'fs h port j j

l p

4

-mo e

~

d CC i.

.a.

e. '

i rr t

NH-(

e Vt '

M-b REACTIVITY CCNTROL SYSTEMS BORATED WATER SOURCE - SHUT 00WN LIMITING CONDITION FOR OPERATION I

3.1.2.5 As a minimum, one of the following borated water sources shall be OPERABLE:

a.

A Boric Acid Storage System and at least one associated Heat Tracing System with:

1)

A minimum contained borated water volume of62 Se+%fts, l

l as feese.,ted in the Core. Opetubs th>Jf; Report.J

-Between-7400-ed '700- :pa-of-bocon, and

2) $ 14 rn nimsm baron conce,1raNon us presealed in & Core Open J

3)

A minimum solution temperature of 65'F.

4 b.

The refueling water storage tank with:

1)

A minimum contained borated water volume -of 20.000 ge44nt, as Pres eded in 46e. Care Operat;ns Lim lh Report' J 2)

A minimum boron concentration of-2000-ppe, and as prese.Jed in N Core opomfln3 Lim lh Rep 3)

A minimum solution temperature of -'1*F.

APPLICABILITh MODES 5 and 6.

ACTION:

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

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

i a.

At least once per 7 days by:

i 1)

Verifying the boron concentration of the water, 2)

Verifying the contained borated water volume, and 3)

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

l b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature when it is the source of borated water and the outside air temoerature is less than 70*F.

l f

McGUIRE - UNITS 1 and 2 3/4I-k1 Amendment No. 80 (Unit 1) j Amendment No. 61 (Unit 2)

.A

b

}% '1 1

REACTIVITY CONTROL SYSTEMS i

BORATED WATER SOURCES - OPERATING i

LIMITING CON 0! TION FOR OPERATION 3.1.2.6 As a minimum, the following borated water source (s) shall be OPERABLE as required by Specification 3.1.2.2:

a.

A Boric Acid Storage System and at least one associated Heat Tracing System with:

1)

A minimum contained borated water volume 4f *0,452 ;;11cas, l

G,;b Report

J as presented i.s ise Co e. Orcesfi9.br:n,and r

2) [-Setween 7000 :nd 7700 ;;:n emeen%1lon aspresented in the Cert O e A m;n ma m hero f

3)

A minimum solution temperature of 65'F.

b.

The refueling water storage tank with:

m. %,,

V i

1)

A contained borated water volume 4f st-leett 372,190-geHow, (

2)

-Betw--

'^^^ nd " ^^ --- ' '----

trA ?")

sen$tk $n fht (L

m'e n boren C tt i Ye'6 sp

' '*H 3)

A minimum solution temperature of 70*F, and l

Wo 4)

A maximum solution temperature of 100*F.

-Wo APPLICABILITY:

MODES 1,_2, 3 and 4.

k3 PY ACTION:

F E

m-a.

With the Boric Acid Storage System inoperabia and being used as one

[. 3 ^

of the above required borated water sources, restore-the storage pp system 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 y

STANOBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and borated to a SHUTDOWN MARGIN 3 P equivalent to at least 1% delta k/k at 200*F; restore the Boric Acid

.,R Storage System to OPERA 8LE status within the next 7 days or be in "g

COLD SHUT 00WN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

J S

k "?

3' b.

With'the refueling water storage tank inoperable,-restore the tank to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STAN08Y

'.ithin the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTOOWN within the following P

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

4.

d

t:

i h

4

'.I O.?

McGUIRE - UNITS 1 and 2 -

3/4 1/712 Amendment No.80_(Unit 1) 7

-Amendment Ho,61-(Unit 2)

. )'

M4 1

i i

f 3/4.5 EMERGENCY CORE COOLING SYSTEMS 4

3/4.5.1 ACCUMULATORS i

COLD LEG INJECTION l

LIMITING CONDITION FOR OPERATION I

3.5.1.1 Each cold leg injection accumulator shall be OPERABLE with:

)

a.

The isolation valve open, b.

A contained borated water volume of between 6870 and 7342 gallons.

i A boron concentration +4 between 1900-end '100 pInI c.

+4s Lcol.uah [ resented h, the Cere. oneensd ba Ar Re$9 psig, and 4

i A nitroge cover pressure of bdtween 58 and 6 j

d.

j e.

A water level and pressure channel OPERABLE.

l APPLICABILITY:

NDES 1, 2, and 3".

1.CO

& lodll,;f peseded in the Cm operatia3 Lim,h Repet ACTION:

i a.

With one accumulator inoperable, except as a result of a closed isolation valve or coron concentration less than 1900 pp ;(* restore the inoperable accumulator-to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in 4

at least HOT STANOBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce Reactor Coolant System pressure to less than 1000 psig within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , a

l b.

With one accumulator inoperable due to the isolation valve being closed, either immediately open the isolation valve or be in at least HOT STANOBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce Reactor Coolant System pressure l

l to lessgan 1000 psig within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

r %e lwer linlf preseded in the Cere. Openfio,q Li lts Repet With one accunfulator inoperable due to boron concentration less than

c. (4900-opa. and:

'y

{

epd Mhe lower limitpresentad i he. Care Oper.+;9 Lin.t,Repet-1)

The volume weighted average boron concentration of the accumula-tors 4900-ppm or greater, restore the inoperable accumulator to 1

OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of the low boron determination i

or be in at least HOT STAN0BY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce Reactor Coolant System pressure to less than 1000 psig within l

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

The volume wefghted average boron concentration of the accumula 9 m,'t>

St. minim.m re%eed to ensure. post-2001 skr%Uy preseJed ls k Care Opndl. fa 4w 2)

Reg.r de leser A!/ preseatW im tors less than,4900 pre but greater than 4800 pp=v restore tus_

' l C,= Cnedl.3 L1~% Ref.et-inoperable accumulator to OPERABLE status or return the volume tco weighted average boron concentration of the -thru li=iting ac-

& /.n#//mit prue.t<J is N _cumulators to greater thag1900 ppe and enter ACTION c.1 within C.ce. opedi 3 4/ !4 Regrr 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or the low baron determination or be in NOT STANOBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce Reactor Coolant System l

pressure to less~than 1000 psig within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , i

1

Reactor Coolant System pressure above 1000 psig.

l i

McGUIRE - UNITS 1 AND 2 3G25 Amendment No. 128(Unit 1)

Amendment No.110(Unit 2)

-. - __.~___ -.

M*0 i

EMERGENCY 20RE C00 LING $YSTEMS LIMITING CON 0! TION FOR OPERATION (Continued) ei c). 'r tors 1400-epe-or less, return the volume weignted average boron he volume weighted average boron concentration of the accumula-ec t 0 4 e3 we C

e'"D concentration of the 4hece--4Mt4*g-accumulatore to greater than i P["'N'",$'r

-le00-opm-and enter ACTION c.2 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of the low boron

)

/cetermination or be in HOT STANOBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and racuce Reactor Coolant System pressure to less than 1000 psig within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ut cepir.d te ensore. put 4004 s Lera./@ pecu ted e n, n ~.,

m the Gre Qpeed.y Li ;1s Repor t SURVEILLANCE REOUIREMENic 4.5.1.1.1 Each cold leg inje eien accumulator shall be demonstrated OPERABLE:

l a.

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

1' 1)

Verifying the contained borated water volume and nitrogen

}

cover pressure in the tanks, and i

2)

Verifying that each cold leg injection accumulator isolation I

valve is open.

b.

At least once per 31 days and within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after each solution volume increase of greater than or equal to 1% of tank volume not t

resulting from normal makeup by verifying the coran concentration 4

of the accumulator solution; l

c.

At least once per 31 days when the RCS pressure is above 2000 psig by verifying that power to the isolation valve operator is disconnected; and d.

At least once per 18 months by verifying proper operation of the power disconnect circuit.

I 4.5.1.1.2 Eacn cold leg injection accumulator water level and pressure channel snall be demonstrated OPERABLE:

1 a.

At least once per 31 days by tl.a performance of an ANALOG CHANNEL OPERATIONAL TEST, and b.

At least once per 18 months by the performance of a CHANNEL 1

CALIBRATION.

1 4

McGUIRE - UNITS 1 AND 2 3/4 5-2 Amendment No.128 (Unit 1)

Amendment No,110 (Unit 2) 2 73

A4 - I C) i EMERGENCY CORE COOLING SYSTEMS 3/4.5.5 REFUELING WATER STORAGE TANK

?

4 LIMITING CONDITION FOR OPERATION

j I

\\.

3.5.5 The refueling water storage tank (RWST) shall be OPERABLE with:

3 a.

A contained borated water volume of at least 372,100 galions, b.

A boron concentration-of between 2000 sad-elOO p 75 af beren, A lits prended ;n Ac Core 0perd;1 *l~F, acd;h RvorT L

A minimum solution temperature of 70 c.

d.

A maximum solution temperature of 100'F.

APPLICABILITY:

MODES 1, 2, 3, and 4.

l ACTION:

With the RWST inoperable, restore the tank to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within 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 .

1 1

i i

SURVEILLANCE RE0VIREMENTS 4.5.5 The r-

' -hall be demonstrated OPERABLE:

a.

At least once per 7 days by:

i 1)

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

Verifying the boron concentration of the water.

A b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by. verifying the RWST temperature when the outside air temperature is either less than 70 F or greater than 100 F.

2-74 4cGUIRE - UNITS 1 and 2 3/4 5-12

A4-ll 1

REACTIVITY CONTROL SYSTEMS BASES 7

^

l * "i MODERATOR TEMPERATURE COEFFICIENT (Continued) 1 tQ hn

'- M and near the end of the fuel cycle are adequate to confirm that the MTCTh remains within its limits since this coefficient changes slowly due principally

.j p to the reduction in RCS boron concentration associated with fuel burnup.

]+. 4 e v1 d 3/4.1.1.4 MINIMUM TEMPERATURN FOR CRITICALITY 5.'!

o E5%

This specification ensures that the reactor will not be made critical 3yE with the Reactor Coolant System average temperature less than 551'F.

This

,, 61 limitation is required to ensure:

(1) the moderator temperature coefficient j o 9 is within it analyzed temperature range, (2) the trip instrumentation is within its normal operating range, (3) the pressurizer is capable of being in e

CfE an OPERABLE status with a steam bubble, and (4) the reactor vessel is above its jt y minimum RT o tS a NDT temperature.

r

-.4 3/4.1.2 BORATION SYSTEMS bbh e 8 The Boron Injection System ensures that negative reactivity control is 3Ig j d'p available during each mode of facility operation.

The components required to

e perform this function include:

(1) borated water sources, (2) charging pumps, i

(3) separate flow paths, (4) boric acid transfer pumps, (5) associated Heat J g's Tracing Systems, and (6) an emergency power supply from OPERABLE diesel generators.

E43 l,o With the RCS average temperature above 200'F, a minimum of two boron

- g _!

injection flow paths are required to ensure single functional capability in j '4E i the event an assumed failure renders one of the flow paths inoperable.

The s1 boration capability of either flow path is sufficient to provide a SHUTDOWN o EN MARGIN from expected operating conditions of 1.3% delta k/k after xenon decay

{

% 4 _+y and cooldown-to 200 F.

The maximum expected boration capability requirement occurs at EOL from full power equilibrium xenon conditions, and requires-y 15,321 galiens-cf 7000 ppe berated u ter frem the beric acid stceage-tanks-oe-600 galicns of 2000 ppe-borated unter from the refueling unter :tcrage tank 4RW6 4 Insed 1 " With the RCS temperature below 200*F, one Boron Injection Syste acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single Boron Injection System becomes inoperable.

The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps except the required OPERABLE pump to be inoperable below 300 F provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV.

McGUIRE - UNITS 1 and 2 B3/41535 Amendment No. 42 (Unit 1)

Amendment No. 23 (Unit 2)

M - 17.,

.'t

.1.

Insert 1 The Technical Specification LCO value for the floric Acid Storage Tank and the Refueling Water Storage Tank minimum contained water volume during Modes 1-4 is based on the the required volume to maintain shutdown margin, an allowance for unusable volume and additional margin as follows:

4 13oric Acid Storage Tank Reauirements For Maintainine SDM - Modes 1-4 Required volume for maintaining SDM presented in the COLR Unusable volume (to maintain full suction pipe) 4,132 gallons Additional margin 6,470 gallons Refueline Water Storace Tank Reauirements t or Maintainine SDM - Modes 1-4 s

Required volume for maintaining SDM presented in the COLR Unusable volume (below nozzle) 16,000 gallons Additional margin 17,893 gallons i

d d

a s

2-76

M 13 1

REACTIVITY CONTROL SYSTEMS i

]

BASES a

BORATION SYSTEMS (Continued)

The boron capability required below 200*F is sufficient to provide a

)

SHUTDOWN MARGIN of 1% delta k/k after xenon decay and cooldown from 200*F to 140*F.

This cond4-tion require: either 2000 gallone of 7000 pp: b:rsted wate-f rc; the -bcHe-ee44-s-torage-t nk: er 10,000 g:llon: Of 2000 ppe berated unte frc; the refueling w:ter ster:ge t:rk.

% serf 2 The contained water volume limits include allowance for water not i

4 available because of discharge line location and other physical characteristics.

The limits on contained water volume and boron concentration of the RWST l

@ z,e,ge also ensure a pH value of between=0,5 erd 10.5 for the solution recirculated

,, 4,i within containment af ter a LOCA.

his pH band minimizes the evolution of iodine and minimizes the effect of hloride and caustic stress corrosion on d e i% i p mechanical systems and components. y, g % J 9. 5 j

1 0#g The OPERABILITY of one Boron Injection System during REFUELING ensures 1

]

O that this system is available for reactivity control while in MODE 6.

,c o g

i

$ [G 3/4.1.3 MOVABLE CONTROL ASSEMBLIES

&a e

9"1 The specifications of this section ensure that:

(1) acceptable power 37 5 distribution limits are maintained, (2) the minimum SHUTDOWN MARGIN is 4

Etv maintained, and (3) the potential effects of rod misalignment on associated

" 1 % accident analyses are limited.

OPERABILITY of the control rod position 1 *h

indicators is required to determine control rod positions and thereby ensure 3 compliance with the control rod alignment and insertion limits.

i o

0 The control rod insertion limit and shutdown rod insertion limits are

! j$e i

specified in the CORE OPERATING LIMITS REPORT per specification 6.9.1.9.

e t

$4 i The ACTION statements which permit limited variations from the basic ggs requirements are accompanied by additional restrictions which ensure that the t

,0 g original design criteria are met.. Misalignment of a rod requires measurement j3 o of peaking factors and a restriction in THERMAL POWER.

These restrictions t

I d provide assurance of fuel rod integrity during continued operation.

?

In 4 e d addition, those safety analyses affected by a misaligned rod are reevaluated I_8 3 to confirm that the results remain valid during future operation.

a s E)3$

The maximum rod drop time restriction is consistent with the assumed rod i"

drop time used in the safety analyses.

Measurement with T greater than or

.1.f p equal to 551*F and with-all reactor coolant pumps operatinP9nsures that the E -4,y'i, measured drop times will be representative of insertion times experienc v.g during a Reactor trip at operating conditions.

$ E cv Control rod positions and OPERABILITY of the rod position indicators are required to be verified on a nominal basis of once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months with more frequent verifications required if an automatic monitoring channel is inoperable.

These verification frequencies are adequate for assuring that the applicable LCO's are satisfied.

C' McGUIRE - UNITS 1 and 2 83/41377 Amendment No.105 (Unit 1) 1 Amendment No. 87 (Unit 2)

/% - 14 l

i l

Insert 2 k

The Technical Specification LCO value for the Boric Acid Storage Tank and the Refueling

)

Water Storage Tank minimum contained water volume during Modes 5 & 6 is based on the the required volume to maintain shutdown margin, an allowance for unusable volume and additional margin as follows:

Boric Acid Storage Tank Reauirements For Maintainine SDM - Modes 5 & 6 Required volume for maintaining SDM presented in the COLR Unusable volume (to maintain full suction pipe) 4,132 gallons Additional margin 1,415 gallons i

3 Refueline Water Storage Tank Reouirements For Maintaining SDM - Modes 5 & 6 l

Required volume for maintaining SDM presented in the COLR Unusable volume (below nozzle) 16,000 gallons 4

Additional margin 6,500 gallons h

0 1

1 l

J 1

2-78

M-l5 4

h 3/4.5 EMERGENCY CORE COOLING SYSTEMS BASES 3/4.5.1 ACCUMULATORS The OPERABILITY of each Reactor Coolant System (RCS) Cold Leg Accumulator ensures that a sufficient volume of borated water will be immediately forced into the reactor core through each of the cold legs in the event the RCS pres-sure falls below the pressure of the accumulators.

This initial surge of water into the core provides the initial cooling mechanism during large RCS pipe rup-tures.

The limits on accumulator volume, bo'ron concentration and pressure ensure that the assumptions used for accumulator injection in the safety analysis are met.

Insect 3

-+ -The - allowed-down-t4me-fee-the-aeeumuletor; cra vari abie-- based-upon-boron-

-eeneenttat4en-to-ensuee-that-the-eeaster i: : hu tdown-feHowi ng-a-WC4-a nd-tha t-

-any problems cre-coreeeted-4n-a-t4mely anner.

Subcriticality i: assured when toren concenteat4en-4+-obove -1500 ppe -so-eddit4er.a1 down time i: 0110wed when r

-concent+et4en-is-above-1500-ppm.

1 c+ncentrat4en-of iesa then-B00-ppe-4n-eny-s4egic accumulator---or-a: c-vele=c weighted-aveeage may-be-indicative of a pro =-

blem, such-as-valve leakage, but since reactae-shutdown-is-essueed --add 4t4enal-T

-tiac is allowed-to-eestoce-boeon-concent+et4en-4n-the-accumulators, The accumulator power operated isolation valves are considered to be

" operating bypasses" in the context of IEEE Std. 279-1971, which requires that bypasses of a protective function be removed automatically whenever permissive conditions are not met.

In addition, as these accumulator isolation valves fail to meet single failure criteria, removal of power to the valves is required.

The limits for operation with an accumulator inoperable for any reason ex-4 cept an isolation valve closed minimizes the time exposure of the plant to a LOCA event occurring concurrent with failure of an additional accumulator which 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.

The Original 'ieensing-bases -cf McGuire assumes-both-the-E-system-and-the Cald Leg Accumulators-function to mitigate postulated ace 4 dent:.

Subse -

quent enalyses, documented in "McGuire Nuclear Stetion, Safety Analysi;-for UliI Climination" dated September 1005, and docketed by Oukc let+ee-det+d

-October 2, 19857-support the-determinatica that UMI is ac icnger required--pro-vided th Cold Leg Aeeumelator vclume is adjusted tc be cons 4: tent "!th that-assumed 4-the Safety Analysis.

T i

'1 McGUIRE - UNITS 1 and 2 B 3/4 6-F Amendment No. 82 -(Unit 1)

Amendment No. 63 (Unit 2)

M-l(i 1

insert 3 The allowed outage time for the accumulators are variable based upon boron ccncentration to ensure that the reactor is shut down following a LOCA and that any problems are corrected in a timely manner. The minimum boron concentration required to ensure post-LOCA suberiticality, as presented in the Core Operating Limits Report, is based on nominal accumulator volume conditions and allows additional outage time since l

suberiticality is assured when the baron concentration is above this value. A slightly higher boron concentration, the minimum accumulator boron concentration limit for LCO 3.5.lc presented in the Core Operating Limits Report, is based on worst case liquid mass, boron concentration and measurement errors. A concentration less than this LCO value in any single accumulator or as a volume weighted average may be indicative of a problem, such as valve leakage. Since reactor shutdown is assured if the boron concentration is above the minimum concentration to ensure post-LOCA suberiticality and the accumulator volume is greater than or equal to the nominal volume, additional time is allowed to restore boron concentration in the accumulators.

1 l

e J

4 e

2-80

/(i-l1 I

EMERGENCY CORE COOLING SYSTEMS BASES REFUELING WATER STORAGE TANK (Continued) for the most reactive control assembly.

These assumptions are consistent with the LOCA analyses.

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

7.5 nJ 9 5 The limits on contained waterfvolume and boron concentration of the RWST also ensure a pH value of between-0.5 and-10.5 for the solution recirculated within containment after a LOCA.

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

i 9

4 a

4 j

McGUIRE - UNITS 1 and 2 8 3/4 38B Amendment No. 82 (Unit 1) l Amendment No. 63 (Unit 2)

. ~.

M-It 4

i k

-I k

i J

4

.i

.I J

1 i

a 4

4 I

l 6

i 7

1 Changes to Core Operating Limits Report (McGuire 1 Cycle 8 & McGuire 2 Cycle 8) 1 1

5 4

4 t

2-82

, =

M-l0 -

e 1

MCE!.0400-05 4of273 9

McGuire l Cycle 8 Cure Operating Limits Report l

l-i 1.0 - Core Ooeratine Limits Reoort 1

}

This Core Operating Limits Report,(COLR), for McGuire. Unit 1, Cycle 8 has been prepared in accordance with the requirements of Technical Specification j

6.9.1.9, a

The Technical Specifications affected by this report are listed below:

.3 Moderator Temperature Coefficie

& bef W Wb I

3/4.1.3.5 tdown Rod insertion 3/4.1.3.6 Contro RmL1 n Limit L3e,.t i j

3/4.2.1

- Axia -

iffere l

3/4.2.2 eat Flux Hot Channel Fact

..3 Nuclear Enthalpy Rise Hot Channel a j

k I

i i

I 1

l i

4-i i

i i

l-I I

f f

s 2-83 e

i

,.. - - +, - -

..,,..r--;

-.,y

,~,,

y

~r v,

,,, U, r 7

rm,

,a

A4 - M Insert 1 2.2.1 Reactor Trip System Instrumentation Setpoints 3/4.1.1.3 Moderator Temperature Coefficient 3/4.1.2.5 Borated Water Source - Shutdown 3/4.1.2.6 Borated Water Source - Operating 3/4.1.3.5 Shutdown Rod Insertion Limit 3/4.1.3.6 Control Rod Insertion Limit 3/4.2.1 Axial Flux Difference 3/4.2.2 Heat Flux Hot Channel Factor 3/4.2.3 Nuclear Enthalpy Rise Hot Channei Factor 3/4.5.1 Accumulators 3/4.5.5 Refueling Water Storage Tank 2-84

A( d )

I MCEl-0400 05 l

$of273 1

l McGuire i Cycle 8 Core Operating Limits Report i

i h

1.1 i

Z0 Doeratinn Limits j

The cycle specific parameter limits for the specifications listed in section 1.0 are presented in the following subsections. These limits have been developed using NRC-approved methodologies specified in Technical Specification 6.9.1.9.

Lw t 2 ---->

i j

Jef Moderator Temocrature Coefficient $necification 3/4.1.1.3)

{

3.0 J:T.I The Moderator Temperature Coefficient (MTC) Limits are:

3.0

{

The MTC.shall be less positive than the limits shown in Figure 1. The BOC/ARO/HZP MTC 3 hall be less positive that 0.7

10E-04 AK/K/ F.

The EOC/ARO/RTP MTC shall be less negative that -4.l* 10E-04 l

AK/K/ F.

Af.2 The MTC Surveillance Limit is:

3.0 l

The 300 PPM /ARO/RTP MTC should be less negative than or equal to I

-3.2

10E-04 AK/K/ F.

t-l Where:

BOC _ stands for Beginning of Cycle ARO stands for All Rods Out.

4 HZP stands for Hot Zero (Thermal) Power EOC stands for End of Cycle l

RTP stands for Rated Thermal Power i

d i '

l 245

- -... - ~,

,.m.

.-.,_,,m.

l

/%. 2.1 Insert 2 1

2.0 Reactor Trio Ssstem Instrumentation Setooints (Snecification 2.2.1) 2,1 Overtemperature AT Setpoint Parameter Values Parameter Value K s 1.1958 Overtemperature AT reactor trip setpoint 1

K = 0.03143/ F Overtemperature AT reactor trip heatup setpoint penalty 2

coefficient K = 0.001405/ psi Overtemperature AT reactor trip depressurization setpoint 3

penalty coefficient a

Measured reactor vessel ATlead/ lag time constants 11 2 8 sec.,

1 t2 5 3 sec.

i Measured AT lay time constant t3 s 2 sec.

Measured reactor vessel average temperature lead / lag I4 2 28 sec, time constants T5 6 4 SCC-Measured reactor vessel average temperature lag time t6 5 2 sec.

constant f (AI)" positive" breakpoint

= 7.07c Al t

f (AI)" negative" breakpoint

= -39.07c Al t

f (AI)" positive" slope

= 1.511% ATo / VcAl t

f (AI) " negative" slope

= 6.153% ATo / 7eAl t

2-86

M - 03 Insert 2 - continued 2.2 Overpower AT Setpoint Parameter Values Parameter Value l

f Overpower AT reactor trip setpoint K = 1.0809 4

K = 0.001239/ F Overpower AT reactor trip heatup serpoint penalty 6

coefficient 4

Measured reactor vessel AT lead / lag time constants t1 2 8 sec.,

t2 5 3 sec.

i Measured ATlag time constant T3 $ 2 sec.

I Measured reactor vessel average temperature lag time 16 s 2 sec.

constant Measured reactor vessel average temperature rate-lag t7 2 5 sec.

l time constant

f (AI)" positive" breakpoint

= 35.07c Al 2

f (AI) " negative" breakpoint

= -35.07c Al 2

f (AI)" positive" slope

= 7.07c ATn / %Al 2

i t

f ( AI) " negative" slope

= 7.0% ATo / 7 cal 2

d 4

f d

9 2-87

-r

.. ~.. -.- - - -.

~.

N.0Y MCE!4M00-05 7 of 273 McGuire I Cycle 8 Cure Operating 1.imits Report i

l Ensect 3 i

2r2' Shutdown Rod insertion I imit (Snecification 3/4.1.3.5) 3.3 26.1 The shutdown rods shall be withdrawn to at least 222 steps.

33 3rJ' Control Rod insertion I.Imits (Snecifiention 3/4.1.3.6) i.

3.'i 2rJ'l The control rod banks shall be limited to physical insertion as shown in M

Figures 2 and :A. Figure 2 applies for s 340 EFPD. Figure 2A applies for l

> 340 EFPD.

i j

Jct' Asial Flux Difference A'occification 3/4.2.1) 3.S 34.1 The AX1AL FLLX DIFFERENCE ( AFD) Limits are provided in Figure 3.

l.

3.3 d

t i

I i

i 4

i i

I t

s 1

l i

j 2 88 i

i

_~.

M.Qs i

i j

1 i

i' l

Insert 3 3.1 Borated Water Source Chutdown (Snecification 3/4.1.2.5 )

i l

3.1,1 Volume and boron concentrations for the Boric Acid Storage System and the Refueling Water Storage Tank (RWST) during modes: 5 & 6:

Parameter Limit i

l Boric Acid Storage System minimum boren 7,000 ppm i

concentration for LCO 3.1.2.5a l

Boric Acid Storage System minimum contained 6,132 gallons water volume for LCO 3.1.2.5a i

l Boric Acid Storage System minimum water volume 585 gallons j

required to maintain SDM at 7,000 ppm l

Refueline Water Storace Tank minimum boron 2,000 ppm concentration for LCO 3.1.2.5b i

l Refueling Water Storage Tank minimum contained ~

26,000 gallons water volume for LCO 3.1.2.5b i

j Refueling Water Storage Tank minimum water 3,500 gallons volume required to maintain SDM at 2,000 ppm i

L 3,2 Hornted Water Source - Oneratine $necifiention 3/4.1.2.6 )

l 3.2.1 - Volume and boron concentrations for the Boric Acid Storage System and the -

l Refueling Water Storage Tank (RWST) during modes: 1,2,3 & 4:

l.

Parameter -

Limit i-

[

Boric Acid Storage System minimum baron _

_ 7,000 ppm concentration for LCO 3.1.2.6a

(

i Boric Acid Storage System minimum contained 20,453 gallons water volume for LCO 3.1.2.6a l

Boric Acid Storage System minimum water volume 9.851 gallons required to maintain SDM at 7,000 ppm l

l 2-89 or

_.m

=

w-u

--,+e,r w -. ie -

M-M insert 3 continued Parameter Limit Refueling Water Storage Tank minimum boron 2,000 ppm concentration for LCO 3.1.2.6b Refueling Water Storage Tank maximum boron 2,100 ppm concentration for LCO 3.1.2.6b Refueling Water Storage Tank minimum contained 91,000 gallons water volume for LCO 3.1.2.6b Refueling Water Storage Tank minimum water 57,107 gallons volume required to maintain SDM at 2,000 ppm l

T P

I l

i f

I t

2-90

(

1

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

... - ~.. ~....... -

&b MCEl-(Mot)-05 11 of 273 McGuire ! C3cle 8 Core Operating Limits Report

llent Flux 1101 Channel Factor.F (X.Y.2)(Snecificatinn 3/4.2.2)

Q 3.6 J.f.1 F

= 2.32 3.G Q'

M2 K(Z)is presided in Figure 4 for Mark BW fuel.

.36

.26.3 K(Z)is prosided in Figure 5 for OFA fuel.

.3. c.

The following parameters are required for core monitoring per the Surveillance Requirements of Specification 3/4.2.2:

.2s4 IFh(X,Y,Z))OP = F0(X,Y,Z)

Mg(X,Y,Z)/(UMT*MT* TILT) 3.6 L

where [Fh(X,Y.Z)]OP = cycle dependent maximum allowable design 1-peaking factor which ensures that the Fn(X,Y,Z) limit will be preserved for operation i.

within the LCO limits (Fh(X,Y,Z) LOP,

[Fh(X,Y,Z)]OP includes allowances for -

calculational and measurement uncertainties.

F0(X,Y,Z) = the design power distribution for F. F0(XsY,Z)is 9

provided in Table 1.

~ M (X,Y,Z) = the margin remaining in core location X,Y,Z to the LOCA g

limit in the transient power distribution. Mg(X,Y,Z) is provided in Table 2..

NOTE:

[ Fh(X,Y.Z > l0P is the parameter identified as Fj^9X,Y.Zi in - DPC-NE-20l l P A.

JK5 (Fh(X,Y ZilRPS = p (x,y,7) * (M (X,Y,Z)/(UMT*MT* TILT))

C 34 2-91

.m,

M -d8 MCE!414(x) 05 i

12 of 273 McGuire i Cycle 8 Core Operating Limits Report where IFh(X.Y.ZHRPS = cycle dependent maximum allowable design peaking factor which ensures that the centerline fuel melt limit will be preserved for operation within the LCO limits. [Fh(X,Y,Z))RPS includes allowances for calculational and measurement uncertainties.

FO(X,Y.Z) = the design power distributions for F. F$(X,Y,Z)is 9

provided in Table 1.

NI (X,Y,Z) = the margin remaining to the CFM limit in core location C

X.Y.Z from the transient power distribution. Mc(X,Y,Z) calculations parallel the Mg(X,Y,Z) calculations described in DPC-NE-201 IPA, except that the LOCA limit is replaced with the CFM limit. M (X,Y,Z)is provided in C

Table 3.

UMT = Measurement Uncertainty (UMT = 1.05).

MT= Engineering Hot Channel Factor (MT = 1.03).

TILT = Peaking penalty that accounts for allowable quadrant power tilt ratio of 1.02.

NOTE: IFh(X,Y,Z)lRPS is the parameter identified as Fy^*(X,Y,Z) in DPC-NE-20llPA.

.266 KSLOPE = 0.078

.3.co where KSLOPE = Adjustment to the K value from OTAT required to i

compensate for each IVe that (Fh(X,Y,Z)]RPS cxceeds it limit.

2-92

MCEl.06)-05 1

258 of 273 i

McGuire ICycle 8 Core Operating Limits Report I

h 1

24' Nuclear Enthalov Rise llot Channel Factor. F n(X.Y.Zi(Soccification 3

i 37 3/4.2.3)

[ Fag (X.Y)]LCO = MARP (X,Y) * [1.0 + (1/RRH) * (1.0 - P)]

25,1 McGuire 1 Cycle 8 Operating Limit Maximum Allowable Radial Peaks.

3,?

(MARP(X,Y)), are provided in Table 4.

The following parameters are required for core monitoring per the Surveillance Requirements of Specification 3/4.2.3:

[ fin (X,Y)]SURV = p (X,Y)

Mag (X,Y)/(UMR

TILT), as identified in DPC-t

}

NE-2011PA.

where 1

l:

UMR = Uncertainty value for measured radial peaks,(UMR = 1.04),

4 TILT = Factor to account for a peaking increase due to an allowable quadrantfiltgtTILT - 1.g ndio of, l.o2- __

F" X,Y) = th"e design power distribution for Fag. F'$i(X,Y)is provided -

Jd2 d

3. 9 l

_in Table 5.

j 2K3 M nlX,Y) = the margin remaining in core location X,Y to the DNB limit 3

l

3. 7 from the transient power distribution. M,sa(X,Y)is provided I-in Table 6.

l' 2dr.4 RRH = 3.34 when 0.0 < P s 1.0, j

3.9 where RRH = Thermal Power reduction seguired to compensate for each 1% that Fag (X,Y) exceeds its limit.

P = Thermal Power l

Rated Thermal Power 26.5 TRH = 0.04 3.'?

(_

2 93 L

M - 30 MCEl-(M(M)-05 259 of 273 McGuire ICycle 8 Core Operating Limits Report

't where TRH = Reduction in OTAT K setpoint required to compensate for t

each 19r that Fan (X,Y) exceeds its limit.

$nsert 4 i

a i

4 I

2-94 8

M 31 1

i

!I Insert 4 I

4 3.8 Accumulators (Soccification 3/4.5.1)

I j

3.8.1 Boron concentration limits during modes: 1,2 and 3:

i j

Parameter Lindts j

Cold Leg Accumulator minimum boron 1,900 ppm l

concentration for LCO 3.5.lc

}

l Cold Leg Accumulator maximum boron 2,100 ppm I

concentration for LCO 3.5.lc l

Minimum Cold Leg Accumulator boron :

1,8'00 ppm concentration required to ensure post-LOCA l

suberiticality if 3.9 Refueline Water Stornee Tank (Soccification 3/4.5.5) 3.9.1 Boron concentration limits during modes: 1,2,3 and 4:

Parameter Limits l

Refueling Water Storage Tank minimum boron -

2,000 ppm i

concentration for LCO 3.5.5b i

Refueling Water Storage Tank maximum boron 2,100 ppm j

concentration for LCO 3.5.5b a

I-

=

1 1

'\\

t k

k*

i i

2-95 i-l

M-32.

MCE!4M4XI-20 4 of 303 McGuire 2 Cycle 8 Core Operating Limits Report 1.0 Core Ooeratine Limits Reoort This Core Operating Limits Report, (COLR), for McGuire, Unit 2. Cycle 8 has been prepared in accordance with the requirements of Technical Specification 6.9.1.9.

The Technical Specifications affected by this report are listed below:

1.3 Moderator Temperature Coefficien kef Sc t wab I

3/4.1.3.

tdown Rod Insertion 3/4.1.3.6 Contro.

Inyn n Limit

^5""

1 3/4.2.1 Axia -

iffert.

3/4.2.2 eat Flux Hot Channel Fac

.3 Nuclear Enthalpy Rise Hot Channel i

2-96

M-33 i

4 Insert 1 2.2.1 Reactor Trip System Instrumentation Setpoints 3/4.1.1.3 Moderator Temperature Coefficient 3/4.1.2.5 Barated Water Source - Shutdown 3/4.1.2.6 Borated Water Source - Operating 3/4.1.3.5 Shutdown Rod Insertion Limit 3/4.1.3.6 Control Rod Insertion Limit 3/4.2.1 Axial Flux Difference 3/4.2.2 Heat Flux Hot Channel Factor 3/4.2.3 Nuclear Enthalpy Rise Hot Channel Factor 3/4.5.1 Accumulators 3/4.5.5 Refueling Water Storage Tank i

a d

2-97

I M.3Y i

1-MCEl-Mn20 l

5 of 303 l

McGuire 2 C.icle M Core Operating Limits Report g

i 1.1 f.0 Oncratine Limits l

The cycle-specific parameter limits for the specifications listed in section 1.0 are l

presented in the following subsections. These limits have been developed using SRC-approved methodologies specified in Technical Specification 6.9.1.9.

Insert 2 4 l

l Af N1oderator Temnerature ('oefficient Goecifiention 3/4.1.1.3) l 30

{_

JW:1 - The Moderator Temperature Coeffi:ient (MTC) Limits are:

I J.0 1

The MTC shall be less positive than the limits shown in Figure 1. The j

BOC/ARO/HZP MTC shall be less positive than 0.7

10E 04 AK/K/ F.

i-i The EOC/ARO/RTP MTC shall be less negative than -4.l* 10E-04 l

AK/K/ F.

i Jer.2 The MTC Surveillance Limit is:

lo The 300 PPM /ARO/RTP MTC should be less negative than or equal to 4

-3.2

10E-04 AKiK/ F.

I Where:

BOC stands for Beginning of Cycle.

ARO stands for All Rods Out

)

HZP stantis for Hot Zero tThermal) Power EOC stands for End of Cycle RTP stands for Rated Thermal Power I

e i

l 2-98 e

v.

+

w

-r we.e w

w.-

v,w-,

y

.w,,-.,,

r va-y u

-'-e--T 9- 'ry-

M.35

\\

j insert 2 2.0 Reactor Trio System Instrumentation Setooints (Soecification 2.2.1) 2.1 Overtemperature AT Setpoint Parameter Values Parameter Value Overtemperature AT reactor trip setpoint K s 1.1958 i

Overtemperature AT reactor trip heatup setpoint penalty K = 0.03143/ F 2

coefficient Overtemperature AT reactor trip depressurization setpoint K = 0.001405/ psi 3

penalty coefficient Measured reactor vessel ATlead/ lag time constants ti 2 8 sec.,

t2 5 3 sec.

Measured AT lag time constant t3 5 2 sec.

Measured reactor vessel average temperature lead / lag t4 2 28 sec.

time constants T5 5 4 sec.

Measured reactor vessel average temperature lag time t6 5 2 sec, constant f (AI)" positive" breakpoint

= 7.0% Al t

f (AI)" negative" breakpoint

= -39.0% Al t

f (AI) "pmitive" slope

= 1.5117e AT / %Al t

o f (AI)" negative" slope

= 6.153% ATo / 7 cal t

2-99

t insert 2 - continued 2.2 Overpower AT Setpoint Parameter Values Parameter Value K = 1.9809 Overpower AT reactor trip setpoint 4

K = 0.001239/ F Overpower AT reactor trip heatup setpoint penalty 6

coefficient hicasured reactor vessel AT lead / lag time constants t[ 2 8 sec.,

T2 s 3 sec.

4 i

hicasured ATlag time constant I3 s 2 sec.

i' N1easured reactor vessel average temperature lag time T6 s 2 sec.

constant hicasured reactor vessel average temperature rate-lag t7 25sec.

time constant S

f (Al)" positive" breakpoint

= 35.0% Al 2

f (AI)" negative" breakpoint

= -35.0% Al 2

f (AI)" positive" slope

= 7.0% ATo / %Al 2

f (AI)" negative" slope

= 7.0% ATn / WAI 2

i e

2-100

~

- - -. ~ _

L M.31 i

i MCE! (M(X).20 lj 7 of 303

,\\1cGuire 2 Cycle 8 Core Operating Limits Report i

l

.Tnsert 3 -+

,26 Shutdown Rod Insertion I.imit (Soec liention 3/4.1.3.5) i i

i 33 s

l 12.1 The shutdow n rods shall be withdrawn to at least 222 steps.

3.3 i

l 2dr Control Rod Insertion I Imits (Snecifiention 3/4.1.3.6)~

3.V

,2rJ.I ~ The control rod banks shall be limited to physicalinsenion as shown in J.V Figures 2 and 2A. Figure 2 applies for s 355 EFPD. Figure 2A applies for

> 355 EFPD.

. g Avini Flux I)ilterence Woecification'3/4.2.1) l

3. 5 l

J.4.1 The AX1 AL FLEX DIFFERENCE (AFD) Limits are provided in Figure 3.

SS' l.

}

i i

l i

i t

)-

1 I

I I-2-101

M.38 Insert 3 3.1 llorated Wnter Source. %utdou n (Soecifiention 3/4.1.2.51 3.1.1 Volume and boron concentrations for the Boric Acid Storage System and the Refueling War-r Storage Tank (RWST) during modes: 5 & 6:

Parameter Limi.1 Borie Acid Storage System minimum boron 7,000 ppm concentration for LCO 3.1.2.5a Dorie Acid Storage System minimum contained 6,132 gallons water volume for LCO 3.1.2.5a Borie Acid Storage System minimum water volume 585 gallons required to maintain SDM at 7,000 ppm Refueling Water Storage Tank minimum boron 2,000 ppm concentration for LCO 3.1.2.5b Refueling Water Storage Tank minimum contained 26,000 gallons water volume for LCO 3.1.2.5b Refueling Water Storage Tank minimum water 3,500 gallons volume required to maintain SDM at 2,000 ppm 3.2 Ilorated Water Source Onerntine (Soccincation 3/4.1.2.6 )

3.2.1 Volume and boron concentrations for the Boric Acid Storage System and the Refueling Water Storage Tank (RWST) during modes: 1,2,3 & 4:

Parameter Limit Borie Acid Storage System minimum boron 7,000 ppm concentration for LCO 3.1.2.6a Doric Acid Storage System minimum contained 20,453 gallons water volume for LCO 3.1.2.6a Borie Acid Storage System minimum water volume 9,851 gallons required to maintain SDM at 7,000 ppm 2 102

Ab 39 I

Insert 3. continised i

i Parameter Limit i

j ltefueling Water Storage Tank minimum baron 2,000 ppm concentration for LCO 3.1.2.6b a

l itefueling Water Storage Tank maximum boron 2,100 ppm d

I concentration for LCO 3.1.2.6b a

)

itefueling Water Storage Tank minimum contained 91,000 gallons water volume for LCO 3.1.2.6b l

Itefueling Water Storage Tank minimum water 57.107 gallons

.i volume required to maintain SDM at 2,000 ppm i

h l

1 i

l l

1 4

l 2-103

._-.._2--_,

M.NO i

j MG14 Moo 20 i

llof 3n)

.\\lc(juire 2 Qcle H Core ()perating 1.imits 1(eport

}

i I

1

h. 5" llent Flux flot Channel l' actor.F(jf N.Y.Z)(Soccification 3]M J,6

)6.1 F

= 2.32 1

J.6 Q

A5'.2 K(Z)is provided in Figure 4 for hiatk BW fuel.

a.6 7

,}5.3 K(Z)is provided in Figure 5 for OFA fuel.

J.6 The following parameters are required for core monitoring per the Surveillance 4

Requirements of Specification 3/4.2.2:

l J4.4 IFh(X.Y.Z))OP = F)X Y,Z)

Nig(X,Y Z)/(Uh1T*h1T* TILT)

.s.c w here (Fht X.Y.Zil0P = cycle dependent maximum allowable design i

peaking factor which ensures that the j

Fn(X Y.Z) limit will be preserved for operation within the LCO limits [Fh(X.Y,Z) LOP.

lit (X.Y,Z))OP includes allowances for calculational and measurement uncertainties.

F0(X,Y.Z) = the design power distribution for Fn. F$(X,Y,Z)is i

provided in Table I for normal operating conditions and in Table 1 A for power escalation during startup operations.

Ntq(X,Y,Z) = the margin remaining in core location X,Y,Z to the LOCA limit in the transient power distribution, hig(X,Y,Z)is provided in Table 2 for normal operating conditions and in Table 2A for power escalation during startup operations.

0P is the parameter identified as F$^*(X,Y,Z) in DPC-NOTE:

IFh(X.Y.Z)l N E-20l l PA.

I 2-1(M

& Y$

MCE14M00 20 12 ol' 301 hicGuire 2 C cle M Core Operating 1.imits Report 3

As.5 II$(X,Y Z)lRPS = F"(X.Y.Z) * (Nic(X,Y,Z)/(Uh1T*h1T* TILT))

34 where ll$(X.Y.7JlRPS = cycle dependent maximum allowable design peaking factor which ensures that the centerline fuel melt limit will be preserved for operation within the LCO limits. [I*n(X,Y,Z)]RPS includes allowances for calculational and measurement uncertainties.

F"(X.Y,Z) = the desigt. power distributions for F. F$(X,Y Z)is g

provided in Table I for normal operating conditions and in Table i A for power escalation during startup operations.

Nic(X,Y Z) = the margin remaining to the CFht limit in core location X,Y 2 from the transient power distribution. hic (X,Y,Z) calculations parallel the hi lX,Y,Z) calculations described g

in DPC SE 20llPA, except that the LOCA limit is replaced with the CFhi limit. hic (X,Y,Z)is provided in Table 3 for normal operating conditions and in Table 3 A for power escalation during startup operations.

Uh1T = Nicasurement Uncertainty (Uh1T = 1.05).

h1T = Engineering lint Channel Factor (MT = 1.03).

TILT = Peaking penalty that accounts for allowable quadrant power tilt ratio of 1.02.

NOTE:

Il$(X,Y.Z)lRPS the parameter identified as F7*(X,Y,Z) in DPC-is NE 20l1PA.

,2di,6 KSLOPE = 0.078 34 2-105 J

MW MCE14Mm-20 J

13 or 303 x

.\\1ctiuire 2 C,5cle N Cure Operating 1.imits Iteport 1

1 l

where KSLOPE =

Adjustment to the K value from OTAT required to i

j compensate for each 1% that [Y (X.Y,Z))RPS exceeds its f

l

limit, 4

i I

l

[

\\

.i j

f l

l t

i 2-106

M3 McEl-(M(n20 16 of 303 i

l Mc(iuire 2 C, sele H Core Operating Limits Report i

,24 kglear ICnthalov Rise flot Channel Factor. F lgf X.Y.Z)(Snecification 3

34 3/4.2.31 4

[F n(X.Y)lLCO = M ARP (X.Y) * [1.0 + (1/RRH) * (1.0 - P)]

4 3

i 3,7 i

.35.1 McGuire 2 Cycle 8 Operating Limit Maximum Allowable Radial Peaks, i

(MARP(X Y)) are provided in Table 4.

1 j

The following parameters are required for core monitoring per the Surveillance Requirements of Specincation 3/4.2.3:

i lF$n(X.YllStJRV = F"gX.Y >

Man (X,Y)/(UMR

TILT), as identified in DPC-N E-20 l l PA.

i where 1

UMR = Uncertainty value for measured radial peaks. (UMR = 1.04).

TILT = Factor to account for a peaking increase due to an allowable quadrant tiltxtTILT - h034. roMo of 10L 3

w p?

FYn(X.Y) = $$$lesign power distribution for F n. F$n(X,Y)is provided c

f.2 3

3 in Table 5 for normal operating conditions and in Table 5A for power escalation during startup operations..

A6.3 M n(X.Y) = the roargin remaining in core location X,Y to the DNB limit 3

3'?

from the transient power distribution. M n(X,Y)is provided 3

in Table 6 for normal operating conditions and in Table 6A for power escalation during startup operations..

)6.4 RRil = 3.34 when 0.0 < P s 1.0,

.3.7 where RRH _= Thermal Power reduction required to compensate for each Ir/c that F n(X,Y) exceeds its limit.

3 P = Thermal Pow er Rated Thermal Power 2-107

&~YY MCE!4)400 20 17 of 303 AlcGuire 2 C cle 8 Core Open ating 1.linits Report 3

.25.5 TRif = 0.04 39 where TRI-l = Reduction in OTAT K setpoint required to compensate for i

each I'7c that F :(X,Y) exceeds its limit.

3i I n.se<+ 4 ->

2-108

M N6 II scri 4 3.8 Accumulators iStecification 3/4.5.1) 3.8.1 Boron concentration limits during modes: 1,2 and 3:

ParameteI Limits Cold Leg Accumulator minimum baron 1,900 ppm concentration for LCO 3.5.lc Cold Leg Accumulator maximum baron 2,100 ppm concentration for LCO 3.5 lc Minimum Cold Leg Accumulator baron 1,800 ppm concentration required to ensure post LOCA suberiticality 3.9 Refueline Water Stornee Tank (Soccification 3/4.5.5) 3.9.1 Boron concentration limits during modes: 1,2,3 and 4:

Parameter Limits Refueling Water Storage Tank minimum boron 2,000 ppm concentration for LCO 3.5.5b Refueling Water Storage Tank maximum boron 2,100 ppm concentration for LCO 3.5.5b 2 109

N'Y&

I Changes to the 14cGuire FSAR I

(

t l

l l

l 2 113 i

N~Y 7.2 Reactor Protection Spiem McGuire Nuclear Station his trip protects the core against low DNI1R and trips the reactor on coincidence as listed in Table 71 with one set cf temperature measurements per loop. The setpoint for this trip is continuously calculated by analog circuitry for each loop by sching the follouing equation:

AT,pmo, - (K - K (1 + T 8) { T 4

1, - 7',n) + K (P - 2235) -JtAQ) y i

2 3

i,,3, 3

C ap)

M whue AT =

Overtemperature AT (Percent of full power AT)

T.,, =

Average reactor coolant temperature ('F)

T *,,, =

Design average reactor coolant temperature at full power ('F)

P=

Pressunzer pressure (psig)

K: =

Setpoint bias (Percent of full power AT)

Ks =

Constant based on the effect of temperature on the DSB limits (Percent of full power AT/'F)

Ka =

Constants based on the effect or pressure on the DSD limits (Percent of full power AT/ psi) ta,ts ts a Time constants (sec)

Laplace transform variable (see )

d s=

bQ=

A function of the neutron flux difference between upper and lower long ion g

chambers. (Refer to Figure 71) (Percent of full power AT)

A separate long ion chamber unit supplies the flux signal for each overtemperature AT trip channel.

Increases in A4 beyond a pre denned deadbr.nd result in a decrease in trip setpoirst. Refer to Figure 7 1.

The required one pressurizer pressure parameter per loop is obtained from separate sensors connected to three pressure taps at the top of the pressurizer. Four pressurizer pressure signals are -

obtained from the three taps by connecting one of the taps to two pressure transmitters. Refer to Section 7.2.2.3.3, " Pressurizer Presswe" on page 7 31 for an analysis of this arrangement.

Figure 71 Sheet 5 shows the logic for overternperature AT trip function. A detailed functional desenption of the process equipment associated with this function is contained in Reference 1 on page 7 36.

b. Overpower AT trip This trip protects against excessive power (fuel rod rating protection) and trips the reactor on coincidence as listed in Table 71, with one set of temperature measurements per loop.He setpoint for each channel is continuously calculated using the follouing equation:

A T, pun, = (Ka - Ks Tan-KdT

- T'3 -f(A4))

u an Where:

Ph 1

AT = Overpower AT (Percent of full power AT) ha4 = A function of the neutron flux difference between upper and lower long ion chamber section. (Percent of full power AT) b6N 2-114 7 14 (01 M AY 1992)

M 48 Technical Justification (McGuire) 9 llfigi

__m.____

__m_. _... _

~ _.

._ _~ -.

h*

i Prcooced Pelecat ion of Certain Itemr from the Technical roecifications to t he COLR Several of too following preposed Technical Specification revisions are relocations of certain iteme. from the Technical Specifications t o t he Cot e (perating Limits Report (COLR).

The justification f or these changes is, to a large extent, common to all of them.

The following paragraphs describe this common justification to avoid repetition.

Technical Justification for Procened Relocations irom Technical Snecificationc to COLR I

Catawba Facility Operating License amendment numberc 74 and 68 dated May 17, 1990 for units 1 and 2 respectively, revised the Catawba Technical Specifications to replace the values of certain cycle-4 specific parameter limits with a reference to the COLR, which contains the values of the limits.

However, additional existing cycle-specific parameter limits in the Catawba Technical Specifications, not included in the above amendments, will have to be revised due to changes in these parameters in support of Unit 2 Cycle 6 operation and reload design.

Similar limits have also changed in recent McGuire fuel cycles.

In addition, McGuire Facility Operating License amendment numbers 105 and 87 dated March 15, 1990 for units 1 and 2 respectively, revised the McGuire Technical Specifications to incorporate an identical COLR methodology.

Therefore, in order to simplify NPC review of identical Technical Specification revision proposals, it is proposed that the McCuire Technical Specifications be changed identically wit h j

respect to the applicable item relocations to the COLR.

In recognition of the burden on licensee and imC resources associated with changes to Technical Specifications, the NRC issued Generic Letter 88-16 on October 4, 1988 encouraging licensees to propose changes to Technical Specifications that are consistent with the guidance provided in the enclosure to the generic letter.

This enclosure provides guidance for the preparation of a license amendment request to modify Technical Specifications that have cycle-specific parameter limits.

With the implementation of this alternative the NRC concluded that reload license amendments for the sole purpose of updating cycle specific parameter limits would be unnecessary.

The proposed revisions described below would relocate the cycle-specific parameter limits from the McGuire Technical Specifications in accordance with the guidance provided in the enclosure to Generic Letter 88-16.

Pronoced Revinion to Technical Onecification Table 2.2-1 It is proposed that the following Technical Specification setpoints be relocated to the Core Operating Limits Report:

Overtemperature AT reactor trip setpoint, K1 e

Overtemperature AT reactor trip heatup setpoint penalty e

coefficient, K2 Over temperature AT reactor trip depressurization setpoint penalty coefficient, K3 Overpower AT reactor trip setpoint, K4 e

Overpower AT reactor trip heatup setpoint penalty coefficient, e

KS e Measured reactor vessel AT lead / lag time constants 11 and T2 10 s ~

=

+

s

<x, e

4

-4 a.

a-4.

aa

_._e

-_.A-.

M u

.+J+._--.e.B..

..-e-M.so i

Measured AT lag time constant T3 e

e Measured reactor vencel avorage t e:rperat ure lead / lag t ime I

constants ta and 15 e

Measured reactor vessel average temperature lag timo constant

{

T6 Measured reactor vessel average temperature rate / lag time e

cons t ant 17 I

e f 1(AI)

positive" breakpoint, i.e.,

the most positive imbalance (axial 21ux difference) at which no overtemperature AT reactor trip setpoint penalty is required due to shewed axial power shapes e

f 1(AI)

negative" breakpoint, i.e.,

the most negative imbalance at which no overtemperature AT reactor trip setpoint penalty is required due to skewed axial power shapes e

11(AI)

positive

slope, i.e.,

the rate at which an overtemperature AT reactor trip setpoint penalty is applied for axial power shapes skewed to the top of the core e

f 3(AI) " negative" slope, i.e.,

the rate at which an overtemperature AT reactor trip setpoint penalty is applied for skewed axial power shapes skewed to the bottom of the core e

f2(AI)

positive' breakpoint, i.e.,

the most positive inbalance (axial flux difference) at which no overpower AT reactor trip setpoint penalty is required due to skewed axial power shapes e

f2(AI)

negative" breakpoint, i.e.,

the most negative imbalance at which no overpower AT reactor trip setpoint penalty is required due to skewed axial power shapes e

f2(AI) " positive" slope, i.e.,

the rate at which an overpower AT reactor trip setpoint penalty is applied for axial power shapes skewed to the top of the core e

f2(AI)

negative" clope, i.e.,

the rate at which an overpower AT reactor trip setpoint penalty is applied for skewed axial power shapes skewed to the bottom of the core The overpower Delta Temperature equation in Note 3 of Table 2.2-1 is I

changed to correct a typographical error Technical Justification The setpoints listed above characterice the Reactor Protection System trip functions which protect the reactor core from departure from nucleate boiling (DNB) and centerline fuel melt (CFM).

The f(AI) i setpoints are calculated by the methodology described in Chapter 4 of Duke Power Company topical report DPC-NE-2011-P-A, " Nuclear Design Methodology for Operating Limits of Westinghouse Reactors *(Reference 2).

This report was approved on January 24, 1990 and is listed in Section 6.9.1.9 of the McGuire Technical Specifications.

Chapter 5 of Duke Power Company topical report DPC-NE-2004, "McGuire and Catawba Nuclear Stations Core Thermal-Hydraulic Methodology Using VIPRE-Ol',

discusses the relationship of the AT reactor trip setpoints to the core safety limits.

This report was approved on November 15, 1991.

As described in Chapter 4 of DPC-ME-20ll-P-A, thermal margin calculations are performed for each fuel cycle to verify that the axial 11

[h - lT h i

flux dif f erence react or trip penalty f unction envelopes, 1 1(AI).uid i

f2(AI), remain consetvative If th" preposei Technical Spoci f i e at inn revision is approved, instances in which this check fails, but for which different envelopes can be demonstrated to both 1) provide adequate protection against licensing basis accidents and 2) be satisfied by the fuel cycle design under consideration, no Technical j

Specification change would be necessary.

i As described in Chapter 2 of Duke power Company topical report-DPC-NE-3001-PA, "McGuire/ Catawba Nuclear Station Multidimencional Peactor Transients and Safety Analysis Physics Parameters Methodology *

(Reference 4), safety analysis physics patameters are checked each 4

cycle to determine that the assumptions of the FSAR Chapter 15 accident analyses remain valid.

If the proposed Technical Specification j

revision is approved, instances in which this check fails for a

{

particular parameter, but for which a different parameter value can be demonstrated to both 1) yield an acceptable result when the applicable i

licensing basis accident (s) are reanalyzed and 2) be satisfied by the j-fuel cycle design under consideration, no Technical Specification change would be necessary.

Reanalysis would be likely, in order to get an acceptable answer, to use revised "K" and

t* values.

Reanalysis in J

this case would also, of course, be performed according to the methodologies approved in DPC-NE-3001-PA and/or Duke Power Company topical report DPC-NE-3002-A, "McGuire/ Catawba Nuclear Station FSAR i

Chapter 15 System Transient Analysis Methodology".

DPC-NE-3001-PA and DPC-NE-3002-A were approved November 15, 1991 and are listed in Section 6.9.1.9 of the McGuire Technical Specifications.

The proposed Technical Specification change is similar to those j

approved on September 16, 1991 as Facility Operating License amendments 191, 191, and 188 for oconee Units 1, 2, and 3, respectively.

In these amendments the Reactor Protection System setpoints for the i

Flux / Flow /Inibalance and Variable Low Reactor Coolant System Pressure trip functions were relocated to the COLR.

As stated in the Oconee Technical Specifications, these trip functions provide equivalent protection against fuel thermal limits, specifically DNB and CFM, as the McGuire overtemperature AT and overpower AT reactor trip functions.

i Recent instances where one or more of these setpoints has been changed in a Technical Specification revision proposal include McGuire 1 Cycle 8, McGuire 2 Cycle 8, Catawba 1 Cycle 7, and Catawba 2 Cycle 5.

i The proposed change to the OPAT equation in Note 3 of Table 2.2-1 is a correction of a typographical error which was likely introduced during I

the McGuire Unit 1 and 2 Technical Specification change submittal for Amendments 131/113 dated 05/7/92.

This change is administrative in

)

nature.

Procosed Revision to Minimum Poron Concentration Limits on ECCS Water Sourcen j

It is proposed that the following minimum boron concentration limits be j

relocated from the Technical Specifications to the COLR:

l The cold leg accumulator minimum boron concentration limit in Technical Specification 3.5.1.1.c i

The Refueling Water Storage Tank (RWST) minimum boron e

concentration limit in Technical Specification 3.5.5.b 12

h<0 i

l l

1 i

1 T._chnical Justification 1

The limits on minimum boron concentration for the accumulators and RWST 1

are verified each cycle to ensure the safety analysis assumptions for j

these parameters remain valid.

The minimum boron concentration limits 1

ensure the reactor will remain subcritical during a Loss of Coolant Accident (Reference 13, Chapter 15.6.5.2 ' Post LOCA Subcriticality i

Evaluation").

The post LOCA subcriticality evaluation provides an available sump mixed mean boron concentration curve that must bound the j

required cold, post LOCA critical boron concentrations for each cycle.

4 The critical boron concentration requirements for each cycle are i

determined using the methodology described in Section 9 of DPC-NF-2010A l

(Reference 14).

In addition, the minimum boron concentration will be evaluated each cycle to ensure the solution recirculated within containment after a LOCA will be maintained at a pH value of s 9.5.

l i

This pH limit minimizes hydrogen production from the corrosion of i

aluminum in containment and reduces the effect of caustic stress corrosion on mechnical systems and components.

The pH band given in the bases for Specification 3/4.5.5 in the McGuire Technical i

Specifications and Specification 3/4.5.4 in the Catawba Technical 3

Specifications is revised to provide consistency between the values at i

both plants while meeting the NRC criteria for sump pH after a LOCA, j

contained in Branch Technical Position MTEB 6-1.

The pH band is

{

tevised to be 2 7.5 and 5 9.5 for both plants.

The ice condenser ice l

bed pH is maintained between 9.0 and 9.5 (Specification 3.6.5.1(a)) and is the most alkaline contributor to the containment sump pH after a LOCA, therefore, the pH will be limitod by the maximum pH limit of the l

ice bed.

The lower limit on the band is chosen such that stress i

corrosion cracking will not occur for an extended period following a l

LOCA and iodine retention in the containment sump water is enhanced.

I A similar Technical Specification revision was approved January 5, 1993 for oconee Units 1, 2, and 3, Amendment Hos., 197, 197, and 194, l

respectively, to move these cycle-specific values to the COLR.

The i

baron concentration value for the Oconee Core Flood Tank (CFT), which perf orms the same function as the McGuire Cold Leg Accumlators (CLA),

was proposed to be relocated to the COUR.

Also, the boron concentration value for the Oconee Borated Water Storage Tank (BWST),

which performs the same function as the McGuire Refueling Water Storage l

Tank (RWST), was proposed to be relocated to the COLR.

{-

Specification 3.5.1.1 Action statement c (2) was revised in the approved

{

McGuire Unit 1 Cycle 8 reload submittal (Reference 9) to base the j

volume weighted average boron concentration on all four accumulators j

instead of just the limiting three.

However, two references to the limiting three accumulators were inadvertently not removed in the i

Technical Specification markups for the McGuire Unit 1 Cycle 8 reloaa j

submittal (Reference 7).. This revision was correctly made and approved in the Catawba Unit 1 Cycle 7 reload submittal (Reference 9).

Therefore, the action statement _is. corrected to remove these-references.

The bases for Specification 3.5.1 includes a paragraph with information regarding the UHI system which should have been removed in Amendment 82 to facility operating license NPF-9 and Amendment 63 to facility operating license UPF-17, dated May 10, 1988.

Since the UFI-system-has i

i 13

..mn..w.

-+.

-,w..

ww-,

nwr we,r-

+-

s - - - -

r

--ww

,-,-+~,--,---mew--

- - ~ *, - - * -

~-

v

- se e --m n e v-

W S3 leen removed at McGuire, the inf ormation is no longer needed and is therefore deleted.

Pror'Oced P.evision to Maximum leren Concenti ation Limit s on L'cc0 Wat er

,courcen It is proposed that the f ollowing inaximum boron concentration limits be telocated from the Technical Specifications to the COLR:

The cold leg accumulator maximum boron concentration limit in Technical Specification 3.5.1.1.c The Refueling Water Storage Tank maximum baron concentration e

limits in Technical Specification 3.5.5.b Technical Justi ficat ien The limits on maximum boron concentration for the accumulators and kWST are based on the minimum boron concentration limits, which are determined using the rnethodology in DPC-NF-2010A (Reference 14), to provide adequate plant operating space between these limits.

They are also evaluated to ensure boron precipitation is precluded following a LOCA (Reference 13, Section 6.3.3).

The boron precipitation analysis uses methods and assumptions described in Westinghouse letter CLC-US-309 dated April 1, 1975 with the principal input paramet. ors given in Table 6-137 of Reference 13.

In addition, the maximum boron concentration will be evaluated each cycle to ensure the solution recirculated within containment after a LOCA will be maintained at a pH value 2 7.5.

This pH limit minimizes the evolution of iodine and reduces t!.e effeet of chloride stress corrosion on rnechnical systems and components.

Pror>00ed Fevisicn to Minimum Volume & Boron Concentration Limits on Dorated Water Sourcec for Reactor Shutdown It is proposed that the following minimum volume and boron concentration limits be relocated from the Technical Specifications to the COLR:

The Boric Acid Storage System minimum volume and boron concentration limits in Technical Specification 3.1.2.5 &

3.1.2.6 l

e The Refueling Water Storage Tank minimum volume and boron concentration liniitc in Technical Specification 3.1.2.5 &

3.1.2.6 i

Technical Justification l

The limits on minimum volume and boron concentration for the Boric Acid 1

Storage System and RWST are verified each cycle to ensure the core shutdown analysis assumptions for these parameters remain valid.

The minimwn volume and bcron concentration limits ensure that negative reactivity control is available during each mode of plant operation.

The minimum volume and boron concentration requirements are based on those t; provide a 1% Ak/k shutdown..ngin f or temperatures less i

14 1

r l

M-k than or equal to 200 *F and 1.3% Ak/k for temperatures greater than 200 "F at all times during a given ryclo.

The required boron r

concentrations are determined using the methodology described in i

Section 4 of DPC-MF-2010A (Reference 14).

Further, the limits on RWST minimun, volume and boron concentration help ensure the reactor will remain subcritical during a Loss of Coolant Accident (Reference 13, Chapter 15.6.5.2 ' Post. LOCA Suberiticality Evaluation").

In addition, the limits on RWST minimum volume and boron concentration are evaluated-each cycle to ensure the solution recirculated within containment af ter a LOCA will be maintained at a pH value 5 9.5.

This pH limit minimi::es hydrogen production f rom the corrosion of aluminum in containment and reduces the effect of caustic stress corrosion on mechnical system.s and components.

The pH band given in the bases for Specification 3/4.1.2 in the McGuire Technical Specifications is revised to provide consintency between the values at both plants while meeting the NRC criteria for sump pH after a LOCA, contained in Branch Technical Position MPEB 6-1.

The pH band is revised to be 2 7.5 and s 9.5 for both plants.

The ice condenser ice bed pH is maintained between 9.0 and 9.5 (Specification 3.6.5.1(a)) and is the most alkaline contributor to the containment sump pH after-a h0CA.

Therefore, the pH will be limited by the maximum pH limit of the ice bed.

The lower limit on the band is chosen such that stress corrosion cracking will not occur for an extended period following a LOCA and iodine retention in the containment sump water is enhanced.

The value for the RWST minimum contained borated water volume prescribed in Specification 3 /4.1.2.6 (b) (1) is based on the ECCS RWST volume requirements in Specification 3/4.5.5 and is a larger-volume than required for reactivitity control, as shown in the bases for Specification 3/4.1,2.

The minimum RWST volume reported in the COLR will be the volume required to maintain reactivity control while Specif ication 3 /4.1.2,6 (b) (1) is changed to refer to the value in Specification 3/4.5.5(a) or the COLR, whichever is larger.

In addition, the McGuire Technical Specification bases for-the allowances assumed in the Boric Acid Storage Tank and RWST minimum contained water volumes are added for clarification.

A similar Technical Specification i

revision was approved January 5, 1993 for Oconee Units 1, 2, and 3, i

Amendment Mos., 197, 197, and 194, respectively, to move these cycle-l specific values to the COLR.

The volume and boron concentration value

)

for the Oconee Concentrated Boric Acid Storage Tank (CBAST), which-performs the same function as the McGuire Boric Acid Tank, was i

relocated to the COLR.

Also, the boron concentration value for the l

Oconee Borated Water Storage Tank (BWST), which performs the same j

function as the McGuire Refueling Water Storage Tank (RWST), was relocated to the COLR.

I l

i o

l l.'

L i

IS l

l 5

.. ~

. ~.

M

]

l t

l j

Poferences J

i 1.

DAW-10159P-A, BWCMV Ccrielaticn of Critical Heat Plux in Mixing I

Vane Grid ruel Assemblies, Babcuck & Wilcox, July 1990.

j 2.

DPC-11E-2 011 P-A, Duke Power Company, 11uclear Design Methodology for Core operating Limits of Westinghouse Reactors, March, 1990.

3.

BAW-10174-A, Mark-BW Reload LOCA Analysis for the Catawba and

]

McGuire Units, Babcock & Wilcox, May 1991.

4.

DPC-NE-3001P, Duke Power Company, Multidimensional Reactor 4

Transients and Safety Analysis Physics Parameters Methodology, Revision 1, November 1991.

l 5.

DPC-NE-2004P-A, Duke Power Company, McGuire and Catawba Nuclear a

Stations Core Thermal-Hydraulic Methodology using VIPRE-01.

j December 1991.

t l

6.

WCAP-10988, Cobra-NC, Analysis for a Main Steamline Break in the Catawba Unit 1 Ice Condenser Containment, Westinghouse Nuclear

.l Energy Systems, November 1985.

I 7.

McGuire !Juclear Station Unit 1, Docket Number 50-369, Cycle 8 Reload Submittal, Duke Power Company, June 26,1991.

l 8.

McGuire Nuclear Station Unit 2, Docket Number 50-370, Cycle 8 2

Reload Submittal, Duke Power Ccopany, December 18,1991.

9.

Catawba Nuclear Station Unit 1 Docket Numbers 50-413 and 50-414, Cycle 7 Reload Submittal, Duke Power Company, April, 13, 1992, j

10.

Catawba Nuclear Station Unit 1, Docket Numbers 50-413 and 50-414, 8

Cycle 6 Reload Submittal, Duke Power Company, January 9, 1991.

j 11.

Catawba Nuclear Station, Final Safety Analysis Report, Ucket i

Nos. 50-413/414.

l 12.

Duke letter to U. S Nuclear Regulato17 Commission, McGuire Nuclear Station Docket Numbers 50-369 and -370 Catawba Nuclear Station Docket Numbers 50-413 and -414 Supplementary Information Relative to Topical Report BAW-10173; Boron Dilution Analysis, Duke Power Company, May 15,1991.

J 13.

McGuire Nuclear Station, Final Safety Analysis Report, Docket j

Nos, 50-369/370.

14.

DPC-11F-2010A, McGuire Naclear Station / Catawba Nuclear Station Nuclear Physics Methodology for Reload Design, Duke Power j

Company, June 1985.

l

^

i 16

&5de Generic Letter 88-16 provides guidance for the removal of cycle-specific parameter limits from the Technical Specifications.

Amendments 105 and 87 dated March 15, 1990 for units 1 and 2 respectively, revised the McGuire Technical Specifications to replace the values of certain cycle-specific parameter limits with a reference to the Core Operating Limits Report (COLR), which contains the values of the limits.

Since approval of this amendment, additional cycle specific parameters have been identified due to revisions which were necessary to support Catawba Unit 2 Cycle 6 operation and past McGuire reloads.

NO SIGNIFICANT HAZARDS EVALUATION

-10 CFR 50.92 states that a proposed amendment involves no significant hazards consideration if operation in accordance with the amendment would not:

1)

Involve a

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

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

Involve a significant reduction in a margin of safety.

This proposed amendment will not increase the probability or consequences of an accident which has been previously evaluated.

The cycle specific parameters which have been identified for relocation to the COLR will be calculated using NRC approved methodology and the Technical Specifications will continue to require operation within the cycle specific parameters.

For the above reasons this amendment is considered administrative, and does not increase the probability or consequences of an accident previously evaluated.

Operation in accordance with this proposed amendment will not create any failure modes not bounded by previously evaluated accidents.

Therefore, this change will not create the possibility of a new or different kind of accident from any kind of accident previously evaluated.

The Technical Specifications will continue to require operation within the bounds of the cycle-specific parameter limits.

The cycle-specific parameter limits will be calculated using NRC approved methodology.

In addition, each future reload will require a 10 CFR 50.59 safety review to assure that operation of the Unit within the cycle-specific limits will not involve a reduction in a margin of safety.

Therefore, no margins of safety are affected by the relocation of cycle-specific parameter limits to the COLR.

Based on the above, Duke has concluded the above that there are no significant hazards -considerations involved in this amendment 4

M-Y request.

The change to Note 3 of Table 2.2-1 corrects a typographical error in the overpower Delta Temperature equation.

This change is administrative in nature, and therefore involves no significant hazards consideration.

The proposed Technical Specification change has been reviewed against the criteria of 10 CFR 51.22 for environmental considerations.

As shown above, the proposed change does not involve any significant hazards consideration, nor increase the types and amounts of effluents that may be released offsite, nor increase the individual or cumulative occupational radiation exposures.

Based on this, the proposed Technical Specification change meets the criteria given in 10 CFR

51. 22 (c) (9 )

for categorical exclusion from the requirement for an Environmental Impact Statement.

l l

i l

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.

ML20127G275

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