ML20095F761

Proposed Tech Specs,Implementing COLR in Accordance W/Gl 88-16 & WCAP-14483

ML20095F761
Person / Time
Site:	Point Beach
Issue date:	12/13/1995
From:	WISCONSIN ELECTRIC POWER CO.
To:
Shared Package
ML20095F756	List: ML20095F752 ML20095F761
References
GL-88-16, NUDOCS 9512190225
Download: ML20095F761 (44)

Text

.-.

2)

Logic Channel A logic channel is a group of relay contact matrices which operate in response to the analog channels signals to generate a protective action signal.

fg.

Instrumentation Surveillance 1)

Channel Check Channel check is a qualitative determination of acceptable operability by observation of channel behavior during operation. This determina-tion shall include comparison of the channel with other independent channels measuring the same variable.

2)

Channel Functional Test A channel functional test consists of injecting a simulated signal into the channel to verify that it is operable, including alarm and/or trip initiating action.

3)

Channel Calibration Channel calibration consists of the adjustment of channel output such that it responds, with acceptable range and accuracy, to known values of the parameter which the channel measures.

Calibration shall encom-pass the entire channel, including equipment action, alarm, or trip, and shall be deemed to include the channel functional test.

.gG.

Shutdown 1)

Hot Shutdown The reactor is in the hot shutdown condition when the reactor is subcritical, by an amount greater than or equal to the margin as specified in Technical Specification 15.3.10 and T,,, is at or greater than 540 F.

9512190225 951213 15.1-3 PDR ADOCK 05000266 P

PDR

4 2)

Cold Shutdown The reactor is in the cold shutdown condition when the reactor has a 1

shutdown margin of at least 1 percent Ak/k and reactor coolant tt:mper-ature is s200'F.

3)

Refueling Shutdown The reactor is in the refueling shutdown condition when the reactor is subcritical by at least 5 percent Ak/k and T., is s140*F. A refueling shutdown refers to a shutdown to move fuel to and from the reactor core.

4)

Shutdown Margin Shutdown margin is the instantaneous amount of reactivity by which the reactor core would be subcritical if all withdrawn control rods were

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tripped into the core but the highest worth withdrawn RCCA remains fully withdrawn.

If the reactor is shut down from a power condition, the hot shutdown temperature should be assumed.

In other cases, no change in temperature should be assumed.

hd.

Power Ooeration The reactor is in power operating condition when the reactor is critical and the average neutron flux of the power range instrumentation indicates greater than 2 percent of rated power.

Q.

Refuelino Ooeration Refueling operation is any operation involving movement of core components (thnse that could affect the reactivity of the core) within the containment when the vessel head is removed.

Q.

Rated Power Rated power is here defined as a steady state reactor core output of 1518.5 MWT.*

k@. Thermal Power Thermal power is defined as the total core heat transferred from the fuel to the coolant.

For Unit 2:

If the Reactor Coolant System raw measured total flow rate is j

<174,000 gpm but 2169,500 gpm, Unit 2 shall be limited to s98% rated power.

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Unit 1 - Amendment No. M5 15.1-4 Neve:bcr 17, 1005 Unit 2 - Amendment No. M9

40.

Reactor Critical l

The reactor is said to be critical when the neutron chain reaction is self-sustaining and k,,, = 1.0.

md, low Power Operation The reactor is in the low-power operating condition when the reactor is critical and the average neutron flux of the power range instrumentation indicates less than or equal to 2% of rated power.-

nN.

Fire Suporession Water System 1

A FIRE SUPPRESSION WATER SYSTEM shall consist of: a water source; pump (s);

and distribution piping with associated sectionalizing control or isolation valves.

Such valves shall include yard post indicating valves and the first valve ahead of the water flow alarm device on each sprinkler, hose standpipe or spray system riser.

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i Unit 1 - Amendment No. 157 15.1-5 Unit 2 - Amendment No. 441 Occcmber 8, 1004 i

1 eg. Dose Eouivalent I-131 Dose Equivalent I-131 shall be that concentration of I-131 (microcurie / gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132. 1-133, I-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites."

pg.

E - Averaae Disintearation Enerav E shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in HeV) for isotopes, other than iodines, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.

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Unit 1 - Amendment No. W 15.1-6 Unit 2 - Amendment No. W Deceder 8, 1004

t 4

15.2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 15.2.1 SAFETY LIMIT, REACTOR CORE Applicability:

Applies to the limiting combinations of thermal power, reactor coolant system pressure and coolant temperatureVreRf6P.*c~o_Tfi_stTsyi.fi~iiE.fl3N_iii_dH{M.-m during A

am operation.

ma Objective:

j To maintain the integrity of the fuel cladding.

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October 5, 1984

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The React'or Protection Sist'em setjoint's a'Fi'desigYed 't6' pfevint;'anyi^ ant'idip'atid

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combination of transient' c'onditions for Reactor Coolant System temperat;ure?

pressure, flow rate, axial flux difference and thermal power level thatjo61d j

result in'a'DNBR,of less \\han the' DNBR limit. L A' tomatic enforcenieht 'bf thiis

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u reactor core safety limits 'is,provided by the'Re' actor Protection System and[Stiis Generator Safety Valves.7 The' react'or cores~afety limits 'represntal'de~iigri s

requirement for establishing the Reactor' Protection, System 'setpoints,j An additional; criterion is ' applied't6the' 0'veFtimp'e'ritori sii'd '0Ver'p'dki'rlD41tiiiT.

setpoints 'of th'e Reactor' Protect 1oii System, to'en's6^rs' that the Reactbr~ Core' Sfety a

Limits are not viblated. The average enthalpy of_the hot leg must be less,than

'or equal to the saturation enthalpy. This ensures that the Delta T used in thess Reactor Protection System setpoints'jsl proportional (to the thermal power l. :

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thersVfety;;analy51s;

~

The React 6r Core' Safety' Limits apply'snlj^'du' ring l'powir 'opsratthh"6ecaQs'e[f tlij the only, time that the reactor is generating significant, thermal p'ower.

s Automa' tic protection functions are r'equiped to be operable ddFing poker'bpeTaffbjj

^

to ensure operation within the Reactor Core SafCty Limits.' TheSteamGeneratod Safety Valves o( automatic protection; actions serve to prevent RCS, heat'~up t6lths ReactorCoreSafetyLimit'sconditionsortoinitiatea:reactortripwhichf6;rces the reactor'into Hot Shutdown.

Setpoints' for'th'e, reactor trip functio,ns aM

~

pecifled,in 15.2.3, Unit 1 - Amendment No.

15.2.1-Unit 2 - Amendment No.

E 4 r, s m a 1C.9 1

1 i s yu s %

4v w 4

4 nEArTAn PADE cAccTV I fufTe n us iv i vi s vvnw wru ui a w aiis e BAYMT O C A F Lt itM f T 1 i

ivans w unw i s viv a i 4

660" N w

650 I

2400 PSIA 640 -

2250 PSIA 650-2000 PSIA Y_

~

o 620-

>e 1775 PSIA 610-600-590-580 8.

.1

.2

.5

.4

.5

.6

.7

.8

.9 1.

1.1 1.2 POWER treaction or nominell Unit 1 - Amendment No. 442 Octcber 27, 1903

6 i

Figure 15.2.1 2 REACT 0P,CCP,E SAFETY LIMIM POINT SEACM 'JNIT 2 670 660 650 2425 psia 640 2250 psia 630 C

e ta 5 620 u

~

m m<i=

i.

e l

610 N

h 600 i

1775 psia 590 580 570 -

I 560 0

0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1

1.1 1.2 Core Power (fraction of 1518.5 MWt)

Unit 2 - Amendment No. M9 November 17, 1905

- - ~. - - - -. -

P l

i l

(3)

Low pressurizer pressure -

21865 psig for operation at 2250 psia i

^

primary s3 stem pressure 21790 psig for operation at 2000 psia primary system pressure.

(4)

Overtemperature 1

AT ( 1 +r,S )

1 1+f'S s;aT, (K,-K,(T( 1 +r,S )-T')( 1 +r,S+K (P-P')-f(41)))

3 where indicated AT at rated power, F

_l AT O

average temperature, "F T

wwyd.anss,.1,.._f

.w 3d33s

.c w

W L.:

.v-.

T' 2'

" ' a c '"

.a_

- - ~

s-"

..i m.

E7. n.n.o

,1l og T, P r

r. n. t. s.

e=

w, pressurizer pressure, psig P

=

ec,sv.<

v-s._pec. i imed., _iuz: met _,h.wgd. OLR.

P' 2235 ^:i-e n.i.

e e n

~..

K 1.30 sp'.efffisdiinithsiCOLR 3

. spec fied init;.==in w(e:x==heICOLR

w=.

K, 0.0200

- = = g:: m w D n gt h e f:: 0{ B erm spe led K

0.000701

=

3 r,

25 00: ipsEiffsdiinithi1COLR

,sen=dn:ar h:efCOLR -

.sLe_c_if._ie : wind. ve 3::

r

=

e e n.,e.

t, a, m, O,n,,,a _ a,m,+

a m. n a m i 4. u..= 1. a m... g {"p"g;_gF{,'{,{

7 o

.gg i

3 w

y.

COLR e, m,, + _ _,,.......m a,m u. t. u..,1. a m.. +.

D,T,,n,

t..a, m, n.

,m,

~

7 o

r,,o, m,n. m,. m _,.

m, m, a,m. 4. u..1. w... n.T. n, "sp"eg^" 'fj'"e"' V "Nh' ' *

dg 4

~

.COLR.

,_m c__.___

-m

_ _. 4 u. u, -. nT,n n.__

v ww, vi vva.m.i.

v, w,.s sw...

n w i

• 'G I ) *,

.' -.' w^, '..-'.*vo-v 'i ". u w^

'a^..^s..w^.'.' '. s w w n - '^'n.-n 'v^^

^

^---~~

Ig w

s

.. w s

w vv y

...2 ua..__.a _. a _. _ m._c u.n.

m,...a_

... _1. a.- m vn

_u-m,

4. u. s,

.....m 4.

_u uv.

vm w.ww.v s vi w yvnw a yw n.w w a s v,i mww e,

i

.. u_

__,__._a u-a __

__...m_a

na_,

4..,, s..m,..___., _ ___,

2..m4, a

yuan.

.v ww su ww.w.

vuswu vu mw. a. i w u

.mwn i w a yvi, w

..a

,iy

..-m.

....,...u. w-m.

,a

.- a

,a,

.m..- u....r.----._-...-,

4,., u... -

.a.

r.....

..r

.w w

v.-...

r w

.r A L a.. m, _

L..a l. u. w,,

a, t,

. L.. a,mm, m.w., a n e. J. u. a l. o,

.-_.A.,a,-+-q,..,.,. 1, a, m.

4, a

a a,

a

.... ~..,,,.

. w wv w r-.

w.,

ws w

__...mm,

4..

__m.

_ c, m,m.a_-...-,,

.,,.._u. u....,..

,.m r_

.. rw -,..,,

r

.n..

i. 4. L 4..

1. ",F,

m a m,m a n,.,.,f, /.A f \\ _ n

/.s i c,a, m,

.E a

a n

.m ru -,

s., - wr s,

y, y,

/L, s.i c, a, m.m. e k, r w m...namaan.

.k...,+.. L. a _ a

4... A n a, t,,a,

,a a,v,.a.n a.A e.

e t, m.

.. w......._,

w

. s.

. L.. wa.A T,

., 4, a c a. n n,4.m...

a.b..a l. l.

... a.... 4,, = 1, 1..u,.,. n J u, a,A.

k, u, kn on m,~a m,-m.,

y ~ rs

.~

w.

w.

w 3 4 u..s 1. w.m..

a t, a.n.

am.

m

-mm-a

. at

.. A yv.mm, t, a m, 11. 4..

1.,

m m. Lu

==

m ai

.i.

s

. v u.

yws wvee.

vi

,u.w.

ww w

v.

v v,7 w,-.. J. u..e l. w...am.

o, r, a.

1 a, f, m,_s. a.A m _a i..a m.

t,,a, m, 1,,1 4.. o.

-mman=+

s.

.. rw. -...

r

.. w f ed m.,.

w.yC.,O...L.,,R spe.,,..c. t.i...v

.~..m-,

n e..t.,h e.,y,w.. -.

n...g:s,r v....,,

u a

w,.

I Unit l'- Amendment No. 4M 15.2.3-2 Oct bcr 28, 1094 i

Unit 2 - Amendment No. M0

)

4

(c) for c:ch percent that the magnitude of q, q, exceed: 17 percent, the AT trip setpcint shall bc automatically reduced by an equivalent of 2.0 percent of rated pcweer (5)

Overpower AT ( ; 11w S) i 3

r*S 1

1

$AI,(X,-K, ( -t,S+1 ) ( 1 +r S ) I-N,[I( 1 w S ) - I ))

where AT indicated AT at rated power, F

=

o T

average temperature, F T'

573.O F (Unit 1) spi {iffEd))jilEsi?0(B E 7. n.. n.or r,. 4. +. o,g T. P rig.

s 1

m w

.si..is._&i.m_iE_d.~li_ffi.f._t.l_iE_TCO_ER f

K.

1.080 cf rated power 0.0252 for increa:ing T s^;is{jfjfdj]Mihi{y0LS j

K.

=

n.n c_ _ a _ _,, 4 _,. v

. v s vi uwws s us s iy a 0.00123 for T 2 T' spjpjfjfd]QMjilt@

K.

=

n.. n. c. _ _. r, y,

fs 10 see s. ji. dYi.sdifn_Rhs..R..O..I.R, 2 sec for Rc cmcat er c^uivalent RTO si.s.Ef..f,fEdDIEM. 6.~TC_O[R Ts

=

n

~

0 000 for Sc;tman er equiv 'ent RTO 2 000 for Rc cmont er equivalcat RTO lsiis{l]fsd3@jsi$0@

r,

=

0 :cc for Sostman er equivalent RTO (6)

Undervoltage - 275 percent of r.ormal voltage (7)

Indicated reactor coolant flow aer loop -

290 percent of normal indicated loop flow (8)

Reactor coolant pump motor breaker open (a)

Low frequency set point 255.0 HZ (b)

Low voltage set point 275 percent of normal voltage.

4 l

Unit 1 - Amendment No. 443 15.2.3-3 October 27, 1003 Unit 2 - Amendment No. 146 t

J

With normal axial power distribution, the reactor trip limit, with allowance for errors"), is always below the @{fjjj core safety limit] c: shown en Figure 15.2.1 1 for Unit I cnd Figur 15.2.1 2 for Unit 2.

If axial peaks are greater than design, as indicated by the difference between top and bottom power range j

nuclear detectors, the reactor trip limit is automatically reduced""".

The overpower, overtemperature and pressurizer pressure system setpoints include i

the effect of reduced system pressure operation (including the effects of fuel densification).

The setpoints will not exceed the core safety limits 0; : hewn in Figur 15.2.1 1 for Unit I cad figur 15.2.1 2 for Unit 2.

i The ovarpower limit criteria is that core power be prevented from reaching a value at which fuel pellet centerline melting would occur. The reactor is prevented from reaching the overpower limit condition by action of the nuclear j

overpower and overpower AT trips.

The high and low pressure reactor trips limit the pressure range in which reactor operation is permitted. The high pressurizer pressure reactor trip setting is lower than the set pressure for the safety valves (2485 psig) such that the reactor is tripped before the safety valves actuate. The low pressurizer pres-sure reactor trip trips the reactor in the unlikely event of a loss-of-coolant accident").

The low flow reactor trip protects the core against DNB in the event of either a decreasing actual measured flow in the loops or a sudden loss of power to one or both reactor coolant pumps. The setpoint specified is consistent with the value used in the accident analysis"). The low loop flow signal is caused by a condi-tion of less than 90 percent flow as measured by the loop flow instrumentation.

The loss of power signal is caused by the reactor coolant pump breaker opening Unit 1 - Amendment No. 443 15.2.3-6 Octcber 27, 1993 Unit 2 - Amendment No. 446

F.

MINIMUM CONDITIONS FOR CRITICALITY Soecification:

1.

Except during lcw power physico tests, the reactor shall-net be made eritical when the-moderatcr temperature cccfficient is more pccitive than 6-is/% lh6MdRptElihjMjli@ii35ffjjjjh}Ephjll3pl@iihlj]lhjjlj{i]j ih!?IfiXElHiiM30W 2.

Reacter power shall not execed 70 pereen-ef Rated Power if-the mcdcrator temperature coefficient i: positive.

32. During an approach to criticality, at least one (1) count per second, attributable to neutrons, shall register on a narrow range source range nuclear instrument.

43.

In no case shall the reactor be made critical (other than for the purpose of low level physics tests) to the left of the reactor core criticality curve presented in Figure 15.3.1-1.

Ak 6l4. The reactor shall be maintained subcritical by at least 1% 7 until normal water level is established in the pressurizer.

Basis:

During the early part of the fuel cycle, the moderator temperature coefficient is calculated to be slightly positive at coolant temperatures below 70 percent of rated thermal power.""')

The moderator coefficient at low temperatures will be most positive at the beginning of life of the fuel cycle, when the boron concentration in the coolant is the greatest.

Later in the life of the fuel cycle, the boron concentrations in the coolant will be lower and the moderator coefficients will be either less positive or will be negative. At all times, the moderator coefficient is negative when 270 percent of rated thermal power.

Suitable physics measurements of moderator coefficient of reactivity will be made as part of the startup program to verify analytic predictions.

Unit 1 - Amendment No. W Unit 2 - Amendment No. W 15.3.1-17 May 3, 1991

4 G.

OPERATIONAL LIMITATIONS i

The following DNB related parameters shall be maintained within th. iimits shewn j M f[ @ ] Q g p)M during Rated Power operation:

1, T,,.'..'.'.'..'...'..'...'b'..-".".'."SF.

i l

2.

Reactor Coolant System (RCS) pressurizer pressure] :h:11 bc ::*r,t:'r.:d:

4_

2.._4._

n.

4.

i.

m e en.c.

.y... 4.. eeen..4,

..i.

mmm yesy

.a sny v..s n

mm.

yas s

j

.J u _ 4.,_.-.-.. 4..

9a.n. n.. 4 j

sinEE.

4_

s,.

y

...s.n

. m yar.

m.<..

yes, k.

11. 4..

s i n. E.E

.J n. 4.._..r.-....

4... Sn.an 4.

4_

m.

r.s, r....

3.

Reactor Coolant System raw measured Total Flow Rate (See Basis).

n.

4..

s.o,..onn.

i m

___ n.

i.

3 r-.

4...

.... m.

u.

n.

4..

i s,,,, n.n.n ___ n.

4..

44

... m.

,r_

Basis:

The reactor coolant system total flow ratel for Uniti 115._dI2.3_%Fi. Of 181,000 p:

~

4s based on an assumed measurement uncertainty of 2.1 percent over thermal design f1ow i,, o n.n.a. ___s.

v. u.

3...

.,n

t. i.

m..,,.... c.. n..d..

s.,.

,rm, 4.

u..;

....__2 4..n

...;.....4.

1 2.,, n.n. a. _, _m_.

v.ud.

y...

se y

nee sa....

is vi ww.m..

m..s...n rs.

..r...

.. s,..y 94,o.i

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

. L.

__.1,

.a. 4. _.. c.1. m.

c. i n, n. a.a.n ___

u. s... n.. _ n. 4..

o, 4,,

.. r s...s.

.i 3r_.

,,,;,n,a................. 4...

m.,.. u..

.,,n..-_

c.,.

,4._.4..

..rr....r....

e f-Lo n., e nn.

v. u.,..

4.

....._,..............___.4...,n. c.

u.;.

,r_.

.a n..a.

.L.a__.1

.A.m. 4 _. fl

m..o r. 1 rc a.n.n....

ft. L.

11. 4..

De.c..m.

9 y....rs.

i

..r im.i wi,n s i.

,ym.

.i

.r n. m n.

s.

m9a n a.A

.a.

1

f. l. am.

4..

l..a,. u..

1.,, A, a.n.n.

3r k.n.

,....1

__m..... k.

.a m.

........i 4.

,4_4..;

.u.__...,

n, n.n ___ -....

4.

n.ou.

... M.

7

. yym,

.y.i..

n is isms...

..a.

.n n.i

.y..

s v....

y..

. c. 4 L.a.A

4. n..L e.

.a t. t. i...... 1 E.. i.1.. P. S. k..

IIIe raw Masureu ii0W 4s TL J (1

.y.

...i.n i

s.

.n ba.:ed upon the use of normalized elbow tap differential pressure which is cal 3brated against a precision flow calorimetric at the beginning of each cycle.

i__

e-

n. 4.

o

,c u._ o... c.

3..

e.g......m.

c,

m...+.

m vs is.

m

.s

.n n

....e

.s n.

e..m m..<.

r s.

...s s, i.,,,A, n. n. n. 3r_

L.n.

s t. e n.. E.n.n.

t.l =.1.. 9 _ L.. 1 1 L 1 4,

4.. A

.s n..ow.

.m.A.oweer

.. m.

3r_,

s

..... r Unit 1 - Amendment No. 465 15.3.1-19 N "::bcr 17, 1005 Unit 2 - Amendment No. M9

15.3.3 EMERGENCY CORE COOLING SYSTEM, AUXILIARY C0OLING SYSTEMS, AIR RECIRCULATION FAN COOLERS, AND CONTAINMENT SPRAY Applicablity:

Applies to the operating status of the Emergency Core Cooling System, Auxiliary Cooling Systems, Air Recirculation Fan Coolers, and Containment Spray.

Ob.iective:

To define those limiting conditions for operation that are necessary:

(1) to remove decay heat from the core in emergency or normal shutdown situations, (2) to remove heat from containment in normal operating and emergency situations, and (3) to remove airborne iodine from the containment atmosphere following a t

postulated Design Basis Accident.

Specification:

A.

Safety Iniection and Residual Heat Removal Systemi 1.

A reactor shall not be made critical, except for low temperature l

physics tests, unless the following conditions associated with that reactor are met:

a.

The refueling water tank contains not less than 275,000 gal. of water with a boron concentration of at le :t 2000 ppm @jl([$

Ihjyyjli?]pej{n{((($1{hj[C0(,8 b.

Each accumulator is pressurized to at least 700 psig and contains at least 1100 ft' but no more than 1136 ft' of water withaboronconcentrationofatle:t2000ppmR@jQhi Mii[tMjpp(jfijdhi((jyil(OI.B. Neither accumulator may be

isolated, c.

Two safety injection pumps are operable, d.

Two residual heat removal pumps are operable, e.

Two residual heat exchangers are operable.

Unit 1 - Amendment No. M Unit 2 - Amendment No. B4 15.3.3-1 Occcmber 1975

~

15.3.8 REFUELING Acolicability:

Applies to operating limitations during refueling operations.

Ob.iective:

To ensure that no incident could occur during refueling operations that would affect public health and safety.

Soecifications:

During refueling operations:

1.

The equipment hatch shall be closed and the personnel locks shall be capable of being closed. A temporary third door on the outside of the personnel lock shall be in place whenever both doors in a personnel lock are open (except for initial core loading).

2.

Radiation levels in fuel handling areas, the containment and spent fuel storage pool shall be monitored continuously.

3.

Core subcritical neutron flux shall be continuously monitored by at least two neutron monitors, each with continuous visual indication in the control room and one with audible indication in the containment available whenever core geometry is being changed. When core geometry is not being changed, at least one neutron flux monitor shall be in service.

4.

Tc. isast one residual heat removal loop shall be in operation. However, if refueling operations are affected by i:le residual heat removal loop flow, the operating residual heat removal loop may be removed from operation for up to one hour per eight hour period.

5.

During reactor vessel head removal and while loading and unloading fuel from the reactor, e-min 4 mum boron concentration of 1800 ppm : hall bc maintained in the primary coolant system )MlDijsilhtilfidj[hlihj@$

M5iNINNAMsR{hiihilC0[R.

6.

Direct communication between the control room and the operating floor of the containmes.t shall be available whenever changes in core geometry are taking place.

Unit 1 - Amendment No. 96 15.3.8-1 September 2, 1985 Unit 2 - Amendment No. 100

3 i

t 15.3.10 CONTROL R0D AND POWER DISTRIBUTION LIMITS Acolicability Applies to the operation of the control rods and to core power distribution i

limits.

Ob.iective To insure (1) core subcriticality after a reactor trip, (2) a limit on potential reactivity insertions from a hypothetical rod cluster control assembly (RCCA) ejection, and (3) an acceptable core power distribution during power operation.

Specification A. Bank Insertion Limits 1.

When the reactor is critical, except for phy:ic tect and control red exercisc, the chutdown bank: chall be fully withdrawn.1 Jhsyhiitd6@

b, m :.7,, = h..

...s

b. _e,,,.... i t.. h..i...,. _,. h:,ey ;.l.~..

.v. ?:. : :.,,,,,,b imi_ts_s.s. Rec. fl..nvd..:

w. -..n,tt.

,...a/s

-.. _a i.y.;.y..:hy v.,;.yn_i.t_e.1.wC..O...L R:.9 7

ank. i.s al. R_

9 e

2

~-

m 2.

When the reactor i; critical, the control bank: : hall be-inserted nc

c.... u. - + u..... + u..,.4..a.. -....,u....

um,. u.. -

3, 4.

...-. c 4,,.., _ _ i. e.. e,. i n

...s y..

Exception: to the in:crtion limit are permitted for phy:ic tc t and

.T..h..,.wg.w:.e:,.y.l.. a;.. v, w.k.m.:,,.:.yh, >,..~ s..>s; z.y,l.:,,t..h.,,,.;i n.w,;t., h.n yl.:.i, a.m.,t,.,v.

....,.:g-e,..;iv..;. m,,i.

s...

. e,.&.c. o.:. t.. r,...;,-,....,.a.n..,s,..i.s...,.a. l..v..

etw :. - t

? n o

=

l contro, rod exerc4+eer__

n n... ;.:.;n.,.;

.spe.,c.t f. iedliE.~th. eKCO.L. O...

~

3.

The shutdown margin shall exceed it: applicable value :: cown in Figure 15.3.10 2 under all Otcady : tate Operating condition from 3602F to full power. An exception to the stuck RCCA component of the shutdown marpn requirement is permitted for physic; tests.

Ss M {ij thel....

~

. =. liiititsgpecifi..dlin.,th,e.,.,1COLRf i

e g

4.

Except for phy ic: te:t:

Chutdown margin of at-4ca:t 1% Ak/k shall be maintained when the reactor c001:nt temperature is le:: then-350 %

5.4.

During any approach to criticality, except for physics tests, the critical rod position shall not be lower than the insertion limit for zero power. That is, if the control rods were withdrawn in normal sequence with no other reactivity change, the reactor would not be critical until the control banks were above the insertion limit.

! Fully withdrawn is defined :: ; bank demand p0:ition equal

+.c or greate than 225 steps.

This definition i Opplicable to shutdown and control hafssr Unit 1 - Amendment No. 454 15.3.10-1 August 25, 100?

Unit 2 - Amendment No. 155

\\

B. Power Distribution limits 1.

a. Except during icw power phy:!c: te:t:, the het channel fact-ees defined in the b;;i: must meet the fclicwing limits: lhjjhiQjh]

sjpj{(@ Q 6 Q @ j lo fcl(@hj]ifjj){$[Qq$$ @l @jl @j])ld s

s.

iii?If31NMi39W

,,,. (2.50).,,,,,

m. x n..g I q\\L Jh n n y s. ;

P

• = s w s - e A A ur n f,_

g(LfAs.VV^n\\L,s n,n.c 2

i a we av -

eN

.1 gA

..re A

e is ng3 r,,si.,v n ti V.s s.

r,;

r "hcre P h the fraction of full powcr at which the core i:

Operating, K(Z) is the function in Figurc 15.3.10 3 nd Z is the ee n -height location of F r o

b. Following a refueling shutdown prior to exceeding 90 percent of rated power and at effective full power monthly intervals thereafter, power distribution maps ! sing the moveable incore detector system shall be made to confirm that the hot channel factor limits are satisfied.

The measured hot channel factors shall be increased in the following way:

(1)

The measurement of total peaking factor, F **, shall be o

ir reased by three percent to account for manufacturing tolerances and further increased by five percent to account for measurement error.

(2)

The measurement of enthalpy rise hot channel factor,F",,

shall be increased by four percent to account for neasurement error,

c. If a measured hot channel factor exceeds the full power limit of Specificaticn 15.3.10.8.1.a, the reactor power and power range i

l h'.gh setpoints shall be reduced until those limits are met.

If subsequent flux mapping cannot, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, demonstrate that the full power hot channel factor limits are met, the overpower l

Unit 1 - Amendmcat No. 4-2-0 15.3.10-2 May 8, 1989 Unit 2 - Amendment No. 443 November 1, 1989

and overtemperature AT trip setpoints shall be similarly reduced and reactor power limited such that Specification 15.3.10.B.1.a above is met.

2.

a.

The indicated axial flux difference (AFD) shall be maintained within the aMewed cperctiencl specc defined by Figure 15.3.10 4 1]iiiilG#JEld!GGiiN0(8 except during physics tests. The physics test exemption applies provided that the thermal power is less than or equal to 85% of Rated Power and the limits of Specification 15.3.10.B.1.a are satisfied.

During suspension of the specification, the thermal power shall be determined to be less than or equal to 85% of rated thermal power at least once per hour.

In addition, the surveillance requirements of 15.3.10.B.1.b shall be performed at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, b.

If the indicated AFD deviates from the Figure 15.3.10 4 limit:

PsMiS@h(j$f315i32{0[BI2[i, the AFD shall be restored to within the Figure 15.3.10 4 limit: Hyliffsih{sipf3MX.3(B12]s within 15 minutes.

If this cannot be accomplished, then reactor power shall be reduced until the AFD is within the envelope or the power level is less than 50 percent of Rated Power.

Normally the rate of power reduction is 15% per hour.

Once AFD has been returned to and maintained within the operating envelope, power level is no longer restricted.

If it is necessary to reduce power to 50%, the Power Range Neutron Flux-High Trip Setpoints shall be reduced to less than or equal to 55 percent within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, c.

A power increase to a level greater than 50 percent of Rated Power is contingent upon the indicated AFD being within the F49ere 15.3.10 4 limits requiremeritM6fil513110iB12?a.

d.

Alarms shall normally be used to indicate non-conformance with the flux difference requirements of 15.3.10.B.2.a and 15.3.10.B.2.b.

If the alarms are totally out of service, the AFD shall be noted and conformance with the limits assessed every hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and half-hourly thereafter.

e.

The indicated AFD shall be considered outside of its limits when at least 2 operable excore channels are indicating the AFD to be outside the limits.

Unit 1 - Amendment No. 86 May 22, 1985 Unit 2 - Amendment No. 90 15.3.10-3 October 5, 1984 i

t t

fLqsis Insertion Limits and Shutdown Marain The reactivity control concept is that reactivity changes accompanying changes in reactor power are compensated by control rod motion.

Reactivity changes associated with xenon, samarium, fuel depletion and large changes in reactor coolant temperature (operating temperature to cold shutdown) are compensated by changes in the soluble baron concentration.

During power operation, the shutdcwn bank: are fully withdr=n.

Fully withdr=n 1 defined ::

bank dem:nd pc: Rica equal to cr greater than 225

tcp:. Thiljhy@pMbjfki{hl@ldj61DjQVjyRjjjjhMh]lhl[CO[R$ Evaluation has shown that positioning control rods at 225 steps, or greater, has a negligible effect on core power distributions and peaking factors.

Due to the low reactivity worth in this region of the core and the fact that, at 225 steps, control rods are only inserted one step into the active fuel region of the core, positioning rods at this position or higher has minimal effect.

This position is varied, based on a predetermined schedule, in order to minimize wear of the guide cards in the guide tubes of the RCCA's.

The control rod insertion limits 'sgE{fjiRiQiMi.s provide for achieving hot shutdown by reactor trip at any time and assume the highest worth control rod remains fully withdrawn. A 10% margin in reactivity worth of the control rods is included to assure meeting the assumptions used in the accident analysis.

So a reactor trip occurring during power operation will put the reactor into the hot shutdown condition, in addition, the insertion limits provide a limit on the maximum inserted rod worth in the unlikely event of a hypothetical rod ejection and provide for acceptable nuclear peaking factors.

The specified control rod insertion limits take into account the effects of fuel densification.

The rods are withdrawn in the sequence of A, B, C, D with overlap between banks. The overlap between successive control banks is provided to compensate for the low differential rod worth near the top and bottom of the core.

WhentheinsertionlimitsspecifiedlinnhelC018areobservedend-thecontrolrod bank: are above the solid line shewn on Figure 15.3.10 1, the shutdown requirement is met.

The maximum shutdown margin requirement occurs at end of core life and is based on the value used in analysis of the hypothetical steam break accident.

Unit 1 - Amendment No. M1 15.3.10-8 August 25, 1994 Unit 2 - Amendment No. M5

F49ere 15.3.10 2 gj}h M j{itXs] Q [Q shows the shutdown margin M equivalent to 2.77% reactivity at end-of-life with respect to an uncontrolled cooldown.

All other accident analyses assume 1% or greater reactivity shutdown margin.

Shutdown margin calculations include the effects of axial power distribution.

One may assume no change in core poisoning due to xenon, samarium or soluble baron.

Power Distribution Design criteria have been chosen which are consistent with the fuel integrity analyses. These relate to fission gas release, pellet temperature and cladding mechanical properties. Also the minimum DNBR in the core must not be less than the limit DNBR in normal operation or in short-term transients.

In addition to the above, the peak linear power density must not exceed the limiting kw/ft values which result from the large break loss-of-coolant accident analysis based upon the ECCS acceptance criteria limit of 2200 F.

This is required to meet the initial conditions assumed for loss-of-coolant accident.

To aid in specifying the limits on power distribution, the following hot channel factors are defined:

F,(Z), Heiaht Dependent Heat Flux Hot Channel Factor, is defined as the local heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods.

Imposed limits pertain to the maximum F (Z) in the o

Core.

Fl, Enaineerina Heat Flux Hot Channel Factor, is defined as the allowance on heat flux required for manufacturing tolerances.

The engineering factor allows for local variations in enrichment, pellet density and diameter, surface area of the fuel rod and eccentricity of the gap between pellet and clad.

Combined statistically, the net effect is a factor of 1.03 to be applied to fuel rod surface heat flux.

Unit 1 - Amendment No. 86 15.3.10-9 ty 22, 1985 Unit 2 - Amendment No. 90 October 5, 1984

i F",, Nuclear Enthalov Rise Hot Channel Factor, is defined as the ratio of the integral of linear power along a fuel rod to the average fuel rod power.

Imposed limits pertain to the maximum F",, in the core, that is the fuel rod with the highest integrated power.

It should be noted that F", is based on an integral and is used as such in the DNB calculations.

Local heat flux is obtained by using hot channel and adjacent channel explicit power shapes which take into account variations in horizontal (x-y) power shapes throughout the core. Thus, the norizontal power shape at the point of maximum heat flux is not necessarily directly related to F",.

For normal operation, it is not necessary to measure these quantities.

Instead it has been determined that, provided the following conditions are observed, the hot channel factor limits will be met:

1.

Control rods in a single bank move together with no individual rod insertion differing by more than 24 steps from the bank demand position, when the bank demand position is between 30 steps and 215 steps. A misalignment of 36 steps is allowed when the bank position.is less than or equal to 30 steps, or, when the bank position is greater than or equal to 215 steps, due to the small worth and consequential effects of an individual rod misalignment.

2.

Control rod banks are sequenced with overlapping banks as described in Figur: 15.3.101})g@0M.

3.

The full-length control bank insertion limits are not violated.

4.

Axial power distribution control procedures, which are given in terms of flux difference control and control bank insertion limits, are observed.

Flux difference refers to the difference in signals between the top and bottom halves of two-section excore neutron detectors.

The flux difference is a measure of the axial offset which is defined as the difference in normalized power between the top and bottom halves of the core.

The permitted relaxation of F", allows radial power shape changes with rod insertion to the insertion limits.

It has been determined that provided the above conditions 1 through 4 are observed, these hot channel factor limits are met.

In Specific:tica 15.3.10.B.I.:, Fg i: crbitr:rily li=ited for p ; 0.5 (except for low p:::r phy:te: te:t:).

Unit 1 - Amendment No. M1-15.3.10-10 Augu:t 25, 1004 Unit 2 - Amendment No, M6

An upper bound envelope of 2.50 times the ncemaltzed peaking factor ax4d dependence of Figure 15.3.10-3 @fcTfjsMEthilC0(RJ consistent with the Technical Specifications on power distribution control as given in Section 15.3.10 was used in the large and small break LOCA analyses. The envelope was determined based on allowable power density distributions at full power restricted to axial flux difference (AI) values consistent with those in Specification 15.3.10.B.2.

The results of the analyses based on this upper bound envelope indicate a peak clad temperature of less than the 2200 F limit. When an F, measurement is taken, both experimental error and manufacturing tolerance must be allowed for.

Five percent is the appropriate allowance for a full core map taken with the moveable incore detector flux mapping system and three percent is the appropriate allowance for manufacturing tolerance.

In the design limit of F",, there is eight percent allowance for uncertainties which means that normal operation of the core is expected to result in a design F"os 1.70/1.08.

The logic behind the larger uncertainty in this case is that (a) normal perturbations in the radial power shape (i.e., rod misalignment) affect F",, in most cases without necessarily affecting F, (b) while the operator has a direct influence on F o through movement of rods, and can limit it to the desired value, he has no direct control over F", and (c) an error in the predictions for radial power shape which may be detected during startup physics tests can be compensated for in F, by tighter axial control; but compensation for F", is less readily available. When a measurement of F" is taken, experimental error must be allowed for and four percent is the appropriate allowance for a full core map taken with the moveable incore detector flux mapping system.

Measurements of the hot channel factors are required as part of startup physics tests, at least each full power month operation, and whenever abnormal power distribution conditions require a reduction of core power to a level based upon measured hot channel factors.

The incore map taken following initial loading provides confirmation of the basic nuclear design bases including proper fuel loading patterns.

The periodic monthly incore mappin] provides additional assurance that the nuclear design bases remain inviolate and identify operational Unit 1 - Amendment No. -lM 15.3.10-11 May 8, 1989 Unit 2 - Amendment No. M3 November 1, 1989

FIGURE 15.3.10-1 GONTROL BANK 4NSERTION-LIMITS PO!NT BEAGN4.!N!TS 1 AND 2 240 I

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Unit 1 A=cndment Mc. 151 Unit 2 Amendment No. 155 August 25, 1994

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~m 15.6.9.2 Uniaue Reoortina Reauirements A.

Intearated leak Rate Test Each integrated leak test shall be the subject of a summary technical report, including results of the local leak rate tests and isolation valve leak rate tests since the last report. The report shall include analysis and interpretations of the results which demonstrate compliance with specified leak rate limits.

B.

Poison Assembly Removal From Soent Fuel Storaae Racks Plans for removal of any poison assemblies from the spent fuel storage racks shall be reported and described at least 14 days prior to the planned activity.

Such report shall describe neutron attenuation testing for any replacement poison assemblies, if applicable, to confirm the presence of boron material.

C.

Overoressure Mitiaatina System Ooeration In the event the overpressure mitigating system (power operated relief valves in the low temperature overpressure protection mode) or residual heat removal system relief valves are operated to relieve a pressure transient which, by licensee's evaluation, could have resulted in an overpressurization incident had the system not been operable, a special report shall be prepared and submitted to the Commission within 30 days.

The report shall i

describe the circumstances initiating the transient, the effect of the system on the transient and any corrective action necessary to prevent recurrence.

Unit 1 - Amendment No. 444 15.6.9-4 August 24, 1989 Unit 2 - Amendment No. 4-24

4 J

l Point Beach Nuclear Plant Core Operating Limits Report Unit 1, Cycle 23 Unit 2, Cycle 22 l

i f

i I

4 i

Note: This report is not part of the PBNP Technical Specifications. This report is referenced in the PBNP Technical Specifications.

___.m.__

TABLE OF CONTENTS 1.0 CORE OPERATING LIMITS REPORT......

.......3 2.0 OPERATING LIMITS......

...4 2.1 Shutdown Margin...

..4 2.2 Moderator Temperature Coefficient..

.4 2.3 Shutdown Bank Insertion Limit.......

.4 2.4 Control Bank Insertion Limits.......

.. 4 2.5 Height Dependent Heat Flux Hot Channel Factor (Fo) and Nuclear Enthalpy Rise Hot Channel Factor (F"m)...............

.5 2.6 Axial Flux Difference................

.......5 2.7 Overtemperatwe AT Setpoint.....

.6 2.8 Overpower AT Setpoint...

.7 2.9 RCS Pressure, Temperature, a.7d Flow Departure From Nucleate Boiling (DNB) Limits......................

.. 7 s

2.10 Accumulator Boron Concentration....

.7

2. I1 Refueling Water Storage Tank (RWST) Boron Concentration..............

.8 2.12 Refueling Boron Concentration..

.8 FIGURE 1: Required Shutdown Margin..

.....9 FIGURE 2: Control Bank Insertion Limits............

.. 10 FIGURE 3: Hot Channel Factor Normalized Operating Envelope.........,.....

..... I 1 FIGURE 4: Flux DifTerence Operating Envelope............

........ 12 TABLE 1: NRC Approved Methodologies for COLR Parameters....

.13 k

l l

Unit 1, Cycle 23 2

SAMPLE Unit 2, Cycle 22

Point Beach Nuclear Plant Core Opcrating Limits Report 1.0 CORE OPERATING LIMITS REPORT This Core Operating Limits Report (COLR) for ' 'oint Beach Nuclear Plant has been prepared in accordance with the requirements of Technical ' pecification (TS) 15.6.9.1.D.

A cross-reference between the COLR sections.nd the PBNP Technical Specifications affected by this report is given below:

COLR Section PBNP TS Inscription 2.1 15.3.10. A.3 S: utdown Margin 2.2 15.3.1.F.1 N aderator Temperature Coefficient 2.3 15.3.10. A.1 S iutdown Bank Insertion Limit 2.4 15.3.10. A.2 C ontrol Bank Insertion Limits 2.5 15.3.10.B. l.a Height Dependent Heat Flux Hot Channel Factor (Fq) and Nuclear Enthalpy Rise Hot Channel Factor (F"m) 2.6 15.3.10.B.2.a Axial Flux Difference 2.7 15.2.3.1.B(4)

Overtemperature AT Setpoint 2.8 15.2.3.1.B(5)

Overpower AT Setpoint 2.9 15.3.1.G RCS Pressure, Temperature, and Flow Departure From Nucleate Boiling (DNB) Limits 2.10 15.3.3. A. l.b Accumulator Baron Concentration 2.11 15.3.3. A. I.a Refueling Water Storage Tank (RWST) Boron Concentration 2.12 15.3.8.5 Refueling Boron Concentration Figure 1 Figure 15.3.10-2 Required Shutdown Margin Figure 2 Figure 15.3.10-1 Control Bank Insertion Limits Figure 3 Figure 15.3.10-3 Hot Channel Factor Normalized Operating Envelope Figure 4 Figure 15.3.10-4 Flux Difference Operating Envelope l

Unit 1, Cycle 23 3

SAMPLE Unit 2, Cycle 22

Point Beach Nuclear Plant Core Operating Limits Report 2.0 OPERATING LIMITS 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 the NRC approved methodologies specified in Technical Specification 15.6.9.1.D.

2.1' Shutdown Margin (TS 15.3.10.A.3) 2.1.1 The shutdown margin shall exceed the applicable val,ue as shown in Figure 1 under all steady-state operating conditions from 350 F to full power. An exception to the stuck RCCA component of the shutdown margin requirement is permitted for physics tests.

2.1.2 Except for physics tests, a shutdown margin of at least 1% Ak/k shall be maintained when the reactor coolant temperature is less than 350 F.

2.2 Moderator Temperature Coefficient (TS 15.3.1.F.1) 2.2.1 Except duriiq low-power physics tests, the reactor shall not be made critical when the moderator temperature coefficient is more positive than 5 pcm/ F.

2.2.2 Reactor power shall not exceed 70 percent of Rated Power if the moderator temper iture coeflicient is positive.

2.3 Shutdown Bank Insertion Limit (TS 15.3.10.A.1) 2.3.1 When the reactor is critical, except for physics tests and control rod exercises, the shutdown banks shall be fully withdrawn. Fully withdrawn is defined as a bank demand position equal to or greater than 225 steps.

2.4 Control Bank Insertion Limits (TS 15.3.10.A.2) 2.4.1 When the reactor is critical, the control banks shall be inserted no further than the limits shown by the lines on Figure 2. Exceptions to the insertion limit are permitted for physics tests and control rod exercises. Fully withdrawn is defined as a bank demand position equal to or greater than 225 steps.

Unit 1, Cycle 23 4

SAMPLE Unit 2, Cycle 22

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Point Beach Nuclear Plant Core Opcrating Limits Report 2.5 IIeight Dependent Heat Flux Hot Channel Factor (Fn) and Nuclear Enthalpy Rise Hot Channel Factor (F%)(TS 15.3.10.B.1.a) 2.5.1 Except during low power physics tests, the Height Dependent Heat Flux Hot Channel Factor must meet the following limits:

Fe(Z)s (2'50) x K(Z) for P>0.5 P

Fo(Z)s 5.00x K(Z) for Ps0.5 where:

P is the fraction of full power at which the core is operating, K(Z)is the function in Figure 3, and Z is the core height location ofFq.

2.5.2 Except during low power physics tests, the Nuclear Enthalpy Rise Hot Channel Factor must meet the following limit:

FE < l.70 x (1+ 0.3(1-P)]

where:

P is the fraction of full power at which the core is operating.

2.6 Axial Flux Difference (TS 15.3.10.B.2.a) 2.6.1 The indicated axial flux difference shall be maintained with'n the allowed i

operational space defined by,ure 4 except during physics tests.

Unit 1, Cycle 23 5

SAMPLE Unit 2, Cycle 22

Point Beach Nuclear Plant Core Operating Limits Report 2.7 Overtemperature AT Setpoint (TS 15.2.3.1.B(4))

Overtemperature AT setpoint parameter values:

T' s

573.9 F (Unit 1) 4 570.0 F (Unit 2) 2235 psig P'

=

Ki l.30 K2 0.0200

=

K3 0.000791

=

25 sec ti

=

3 see T2

=

t3 2 sec for Rosemont or equivalent RTD

=

0 sec for Sostman or equivalent RTD

=

2 sec for Rosemont or equivalent RTD T4

=

0 sec for Sostman or equivalent RTD

=

l f(AI) is an even function of the indicated difference between top and bottom detectors of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests, where qi and qs are the percent power in the top and bottom halves of the core respectively, and qi + qs is total core power in percent of rated power, such that:

(a) for gi - qs within -17, +5 percent, f(AI) = 0.

(b) for each percent that the magnitude of gi-qs exceeds +5 percent, the AT trip setpoint shall be automatically reduced by an equivalent of 2.0 percent of rated power for Unit 1, or by an equivalent of 3.1 percent of rated power for Unit 2.

i (c) for each percent that the magnitude of gi-qs exceeds -17 percent, the AT trip setpoint shall be automatically reduced by an equivalent of 2.0 percent of rated power.

Unit 1, Cycle 23 6

SAMPLE Unit 2, Cycle 22

Point Beach Nuclear Plant Core Operating Limits Report 2.8 Overpower AT Setpoint (TS 15.2.3.1.B(5))

Overpower AT setpoint parameter values:

T' 5

573.9 F (Unit 1) s 570.0 F (Unit 2)

L 1.089 0f rated power 0.0262 for increasing T K5

=

0.0 for decreasing T

=

0.00123 for T 2 T' L

=

0.0 for T < T'

=

10 see T5

=

T3 2 sec for Rosemont or equivalent RTD

=

0 sec for Sostman or equivalent RTD

=

T4 2 sec for Rosemont or equivalent RTD

=

0 sec for Sostman or equivalent RTD

=

2.9 RCS Pressure, Temperature, and Flow Departure From Nucleate Boiling (DNB)

Limits (TS 15.3.1.G) 2.9.1 T.y, shall be maintaas i helow 578 F 2.9.2 Reactor Coolant System (RCS) pressurizer pressure shall be maintained:

a.

Unit 1:

21955 psig during operation at 2000 psia.

b.

Unit 2:

21955 psig during operation at 2000 psia.

2.9.3 Reactor Coolant System raw measured Total Flow Rate shall be maintained:

a.

Unit 1:

2181,800 gpm b.

Unit 2:

2174,000 gpm The RCS flow rate limit for Unit 2 at rated power is 174,000 gpm.

However, Unit 2 is analyzed to support operation with a reactor coolant system total flow rate limit of 169,500 gpm. This is based on an assumed measurement uncertainty of 2.1 percent over a thermal design flow of 166,000 gpm. If the Unit 2 RCS raw measured total flow rate is less than 174,000 gpm but greater than or equal to 169,500 gpm, operation is limited to less than or equal to 98% rated power.

2.10 Accumulator Boron Concentration (TS 15.3.3.A.1.b) 2.10.1 The accumulator boron concentration shall be at least 2000 ppm.

Unit 1, Cycle 23 7

SAMPLE Unit 2, Cycle 22

Point Beach Nuclear Plant Core Operating Limits Report 2.11 Refuciing Water Storage Tank (RWST) Boron Concentration (TS 15.3.3.A.1.a) 2.11.1 The refueling water storage tank baron concentration shall be at least 2000 ppm.

2.12 Refueling Boron Concentration (TS 15.3.8.5) 2.12.1 During reactor vessel head removal and while loading and unloading fuel from the reactor, a minimum boron concentration of 1800 ppm shall be maintained in the primary coolant system.

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Unit 1, Cycle 23 8

SAMPLE Unit 2, Cycle 22

Point Beach Nuclear Plant Core Operating Limits Report Figure 1: Required Shutdown Margin s

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aw u

=

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10 20 30 40

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70 00 00 100 s wcutevanw i

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Unit 1, Cycle 23 9

SAMPLE Unit 2, Cycle 22

Point Beach Nuclear Plant Core Operating Limits Report Figure 2: Control Bank Insertion Limits too

!21%}

ISS.so%I 125 4

m 21o 195 Bank 3 inserton l g

/

l185 1g5 176!

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Power Level l% of Rated Peerer) 4 Note:

The " fully withdrawn" parking position range can be used without violating this Figure.

4 Unit 1, Cycle 23 10 SAMPLE Unit 2, Cycle 22

Point Beach Nucle:r Plant Core Operating Limits Report Figure 3: Hot Channel Factor Normalized Operating Envelope 1.2 (0 0.1.0) I (s.o,1.0 ) l 3,3 1.

(12.0. 92)1 e

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Unit 1, Cycle 23 11 SAMPLE Unit 2, Cycle 22

Point Beach Nuclear Plant Core Operating Limits Report Figure 4: Flux Difference Operating Env ' ope I

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Unit 1, Cycle 23 12 SAMPLE Unit 2, Cycle 22 i

o Point Beach Nuclear Plant Core Operating Limits Report Table 1: NRC Approved Methodologies for COLR Parameters COLR Parameter NRC Approved Methodology Section 2.1 Shutdown Margin WCAP-9273-NP-A, " Westinghouse Reload Safety Evaluation Methodology," July 1985 2.2 Mourator Temperature Coeflicient WCAP-9273-NP-A, " Westinghouse Reload Safety Evaluation Methodology," July 1985 2.3 Shutdowa Bank Insertion Limit WCAP-9273-NP-A, " Westinghouse Reload Safety Evaluation Methodology," July 1985 2.4 Control Bank Insertion Limits WCAP-9273-NP-A," Westinghouse Reload Safety Evaluation Methodology," July 1985 2.5 Height Dependent Heat Flux Hot WCAP-9273-NP-A, " Westinghouse Reload Channel Factor (Fn) and Nuclear Safety Evaluation Methodology," July 1985 Enthalpy Rise Hot Channel Factor (F%)

Thadani to Johnson, " Acceptance for Reference of Licensing Topical Report, WCAP-10924, ' Westinghouse Large Break LOCA Best Estimate Methodology,'

Addendum 4, 'Model Revisions'," February 1991 2A

' ' l Flux DifTerence WCAP-9273-NP-A," Westinghouse Reload Safety Evaluation Methodology," July 1985_

2.7 Overtemperature AT Setpoint WCAP-9273-NP-A, " Westinghouse Reload Safety Evaluation Methodology," July 1985 2.8 Overpower AT Setpoint WCAP-9273-NP-A, " Westinghouse Reload Safety Evaluation Methodology," July 1985 2.9 RCS Pressure, Temperature, and Flow WCAP-9273-NP-A, " Westinghouse Reload Departure From Nucleate Boiling Safety Evaluation Methodology," July 1985 (DNB) Limits WCAP-11397-P-A, " Revised Thermal Design Procedure," April 1989 2.10 Accumulator Boron Concentration WCAP-9273-NP-A, "Westnghouse Reload Safety Evaluation Methodology," July 1985 2.11 Refueling Water Storage Tank (RWST)

WCAP-9273-NP-A, "Westinghoux Reload Baron Concentration Safety Evaluation Methodology," July 1985 2.12 Refueling Boron Concentration WCAP-9273-NP-A, " Westinghouse Rehad Safety Evaluation Methodology," July 19b5 l

Unit 1, Cycle 23 13 SAMPLE Unit 2, Cycle 22