Text

Proposed Tech Specs,Allowing Operation of Core within Axial Flux Difference Band Expanded from Current Band in TSs & Improved Fuel Cycle Mgt Through Implementation of Core Design Enhancements
	ML20058C316
Person / Time
Site:	Seabrook
Issue date:	11/23/1993
From:	; NORTH ATLANTIC ENERGY SERVICE CORP. (NAESCO)
To:	;
Shared Package
ML20058C283	List: ML20058C279; ML20058C299; ML20058C316; ML20058C338; ML20058C358;
References
	NUDOCS 9312020417
	Download: ML20058C316 (100)
	v • d • e

_ _ _ _

DEFINITIONS CONTAINMENT INTEGRITY

- - 1.7 +CO'NTI5NMENT INTEGRITY shall exist when:

~ - - ~ "

~'-

^~

+.:.;; -.. u.--

=-

a.

,All penetrations required to be closed during accident. conditions _. ~u.

are either:

2--

1)

Caoable of being closed by an OPERABLE containment automatic isolation valve system, or 2)

Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions.

j b.

All equipment hatches are closed and sealed, c.

Each air lock is in compliance with the requirements of Specification l

3.6.1.3, i

I d.

The containment leakage rates are with'.i the limits of Specifica' tion 3.6.1.2, and The sealing mechanism associated with each penetration (e.g., welds, e.

bellows, or 0-rings) is OPERABLE.

l CONTROLLED LEAKAGE l

l 1.8 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals.

f CORE ALTERATION

1. 9 CORE ALTERATION shall be the movement or manipulation of any component I

l within the reactor pressure vessel with the vessel head removed and fuel in the vessel.

Suspension of CORE ALTERATION shall not preclude completion of j

movement of a component to a safe conservative position.

CORE OPERATING LIMITS REPORT 1.10 The CORE OPERATING LIMITS REPORT provides core operating limits for the 4

l current operating reload cycle. The cycle specific core operating ifaits shall i

be determined for each reload cycle in accordance with Specification 6.8.1.6.

Plant operation within these operating limits is addressed in individual l

speci,fications.

DIGITAL CHANNEL OPERATIONAL TEST 1.11 A DIGITAU CHANNEL OPERATIONAL TEST shall consist of exercising the digital computer hardware using data base manipulation and/or injecting simulatedTheDigital process data to verify OPERASILITY of alam and/cr trip functions.

Channel Operational Test definition is only applicable to the Radiation Monitoring Equipment.

O Amendment No. 9 SEABROOK - UNIT 1 1-2 9312020417 931123 a

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4 0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 1.1 1.2 FRACTION OF RATED THERMAL POWER FIGURE 2.1-1 REACTOR CORE SAFET( LIMIT - FOUR LOOPS IN OPERATION

(

TABLE 2.2-1 REACTOR TRIP SYSTEN INSTRUMENTATION TRIP SETPOINTS SENSOR TOTAL ERROR X

FUNCTIONAL UNIT ALLOWANCE (TA)

Z (S)

TRIP SETPOINT ALLOWABLE VALUE 1.

Manual Reactor Trip H.A.

N.A.

H 2.

Power Range, Neutron Flux a.

liigh Setpoint 7.5 4.56 0

1109% of RTP*

1111.1% of RTP*

b.

Low Setpoint 8.3 4.56 0

125% of RTP*

127.1% of RTP*

3.

Power Range, Neutron Flux, 1.6 0.5 0

<5% of RTP* with

<6.3% of RTP* with i time constant i time constant liigh Positive Rate 12 seconds

>2 seconds 4.

Power Range, Neutron Flux, 1.6 0.5 0

<5% of RTP* with

<6.3% of RTP* v.ith i time constant a time constant y

liigh Negative Rate 22 seconds 32 seconds

+

5.

Intermediate Range, 17.0 8.41 0

125% of RTP*

$31.1% of RTP*

Neutron Flux 5

5 6.

Source Range, Neutron Flux 17.0 10.01 0

$10 cps

$1.6 x 10 cps 7.

Overtemperature AT See Note 1 See Note 2 f

8.

Overpower AT 4-7 See Note 3 See Note 4 9.

Pressurizer Pressure - Low

@ M 11945 psig 1

Pressurizer Pressure - liigh 1-Ge 9-99 12385psig

h. 10.

RTP = RATED THERMAL POWER 1.7 an he sens error f Pressuri Pressure s 0.5.

measured' r T,yg

""Ihe sor tror rse or e ors a be us in lieu f either r both o these val

, which t n must be ummed to ter-

'et over eratu AT tot channel alue for

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TABLE 2.2-1 (Continued) m h

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS

=0 SENSOR 7

TOTAL ERROR FUNCTIONAL UNIT ALLOWANCE (TA)

Z (S)

TRIP SETPOINT ALLOWARIE VALUE c.

=

11.

Pressurizer Water Level - High 8.0 4.20 0.84

<92% of instrument

<93.75% of instrument P"

E*" crneasurM (me454re

>90% of loo

>89.3%offloop l

@ [flo 12.

Reactor Coolant Flow - Low 2.5 1.9 0.6

% f1066 l

13. Steam Generator Water 14.0 12.53 0.55

>14.0% of narrow

>12.6% of narrow Fange instrument range instrument Level Low - Low span span

14. Undervoltage - Reactor 15.0 1.39 0

>10,200 volts

>9,822 volts Coolant Pumps 15.

Underfrequency - Reactor 2.9 0

0

>55.5 llz

>55.3 Ilz Coolant Pumps 16.

Turbine Trip a.

Low Fluid Oil Pressure N.A.

N.A.

>500 psig

>450 psig~

b.

Turbine Stop Valve N.A.

N.A.

>1% open

>1% open i Closure

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SafetyInjectioninput N.A.

N.A.

H.A.

N.A.

from ESF y

E.

W W

A

TABLE 2.2-1 (Continued)

(A9 TABLE NOTATIONS E

8 NOTE 1: OVERTEMPERATURE AT

^

l (ihy,5)

T'] + Ka(p - P') - f (Al))

[T y,,

(7 gsg) $ AT, Mi - K AT i

2 1+

I z4 Heasured AT by RTD Instrumentation; Where:

AT

=

Lead-lag compensator on measured AT; l

f{

=

l Timeconstantsutilizedinlead-lagcompensatorforAT,ri>8s,!

=

T,T2 1

12 $ 3 Si yfg3 Lag compensator on measured AT;

=

a ta Time constants utilized in the lag compensator for AT, 13 = 0 s;

=

Indicated AT at RATED TilERMAL POWER; AT,

=

K

=

olue speci ficci,#T'O L R -

l i

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= grGH2/8@

f

= The function generated by the lead-lag compensator for T,yg dynamic compensation; g

h T4. Ts

= Time constants utilized in the lead-lag compensator for T,yg, 1

> 33 s, 4

a Ts 14 s; 5

Average temperature, "F;

)

T

=

2O g

3 Lag compensator on measured T,yg;

=

Ta Time constant utilized in the measured T,yg lag compensator, is = 0 s;

=

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O- _ _ -

TABLE 2.2-1 (Continued) m h

TABLE NOTATIONS Bg NOTE 1:

(Continued)

T'

$ 588.5"f (Nominal T at RA7ED TilERMAL POWER);

e avg f eci f-iecl i2 C 0L R)

Ka

= M 519/pti

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Pressurizer pressure, psig; l

P

=

2235 psig (Nominal RCS operating pressure);

P'

=

i Laplace transform operator, s 8; M 3 3 p g ] 4;g g ; y gl,, Q S

=

and f (AI) is a function of the hndicated dif ference between top and bottom detectors of the i

onwer-rante neutron 100 chambe"is witn gai to be ectea be a on mea Ped instr ent res nse cur g plant tartup iEsts so t:

g gb ween 5% and

,f(

= 0, wh qt and are perc RATED RMAL ry

) For ER in e top d botto alves o he core r pectively and qt

b is tot TilERMA POWER per t of RAT THERMAL WER; l

(2) F each reent th the mag ude of q gb excee

- 35%,

AT Tri Setpoin shall e au atically educed b

.09% of I value at TED Tile L POWER and 1

y

3) F each per nt that e magnitud of qt exceeds +

, the /

Trip Se oint s 1

l e aut cally re ced by 1.0 of its v se at RAT TilERMA OWER. f a

c^

uted IM Setsei-* by ::: r: than2.5%)

p{ghan.el': =*4:= Tr4p-SHpeint 05:" n:t ext::f It g

NOTE 2:

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LLA OZ SEABROOK - UNIT 1 2-9

TABLE 2.2-1 (Continued) m TABLE NOTATIONS (Continued) is a

7 NOTE 3:

(Continued)

\\/o.9,u e s ecified ib Col,R E

K.

= -0:081386/8f fer ! >.

nd-6 = 0- for T i IS

-i As defined in Note 1, e

T

=

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

=

instrumentation, 5 588.5'F),

As defined in Hote 1, and S

=

Cfv(AI) fee-aH-A D rf I NOTE 4:

he-channel 4-maximum-Te4p 5:tpoint-shal4-neb-exceed-Its cerpeted-ir-sp-setpetat by eere tnag g

r0%-efATspag

\\

g e..h d ey20ent u a.0.u?f ay the c.ha.nn e k 's ABn.sh. g% e.

is spe cs-fj ed in kO L R_f

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(AT) k Nt<neHon d Oe.'incAtcrieb Aibler enbefurs.a.

7 m a ww s.,w+ r, 4 h per rany m.4 ion e hrm b ers as speeilled. inytOLK.

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2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation that would result in the release of fission Overheating of the fuel cladding is prevented

}

products to the reactor coolant.

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

Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure j

from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer DNB is not a directly measurable parameter during operation and, i

coefficient.

therefore, THERMAL POWER and reactor coolant temperature and pressure have been l

M x-! c " c9 related to DNB.OrxqrO 2 ' ~-Cr' a ' -- r ! :ti er l

-rc m has been ceveloped to predict the DNB flux and the location of DNB for axially The local DNB heat flux f

uniform and nonuniform heat flux distributions.

i ratio (DNBR) is defined as the ratio of the heat flux that would cause DNB at

'j a particular core location to the local heat flux and is indicative of the

//

margin to DNB.

gu g a g 7, q

o

~ -state oper ion, norm j

u imum val of the 0' during ste y

g ene m This transie

.s, and an+ cicated trar.ents is lim' ed to 1.30 Nalue c responds a 95% pr ability at 95% confiden level tha DNB wil' f

, ;5peratior accrocriate..arnin to DND for all coe ti n h' not oc r and is nosen as condi ions.f The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System oressure, and average temperature for which the minimum or the average enthalpy' at the vessel exit is equal DNBR is no less tha to the enthalpy of turated licuid. _

4.,;f %

he scde ry a.n d g.s s D N B w

h, of

-~

.55

..mnoei Tactor, a

se arve e=

cu vu a c....

and ref renc cosin with -

eak o

1. 55 ' r axi al ower s pe.

A allow ce is inc1 ded r an creas in F t red ed powe based the e ressi N

N

.55 [1 0.2 (' P)]

F

=

Whe P is, e fr

. ion RATED T ' Py.AL P

'ER.

These mitin heat ux co. itions a high than iose cula d te th range all ntrol ods f' ly with awn to he ma mum a owabl con el d inser on, a uming he i power mbalan is w' hin t limi of he (aI) nctio of th Over+ mperatur trip.

' hen t axi powe imb ance is not ithin

.e tol ance the axi power balan eff t on e0 r-1 temp ture - trips ill duce th Setpoi s to vide rote ion onsist.t witr core S ety Li its.f y

Se. :t r,m t " 6. 2 T/. /. O 1

Sl SEABROOK - UNIT 1 B 2-1

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Insert "B.2.1.1.a" The ONB design basis is as follows: uncertainties in the DNBR correlation, plant operating parameters, nuclear and thermal parameters, fuel fabrication parameters, and computer codes are considered statistically such i

there is at least a 95 percent probability with 95 percent confidence level that ONB will not occur en the most limiting fuel rod during Condition I and II events. This establishes a design DNER value which must be met in plant safety analyses using values of input parameters without uncertainties.

In addition, margin has been maintained in the design by meeting safety analysis CNER limits in performing safety analyses.

Insert "B.2.1.1.b" These curves are based en an enthalpy rise hot channel factor, 01, at RATED THERMAL POWER of 1.65.

The value of 01 at reduced power is assumed to vary according to the expression:

E& = 1.65 [l+ 0.3 (1-P)]

Where P is the fraction of RATED THERMAL POWER.

g This expression conservatively bounds the cycle specific limits on 05 specified in Technical Specification 3/4.2.3 and COLR.

The safety limits in Figure 2.1-1 are also based on a reference cosine axial power shape with a peak of 1.55.

The resulting heat flux conditions are more limiting than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion, assuming the axial power imbalance is within the limits of the f (AI) and f (AI) functions of the Overtemperature and 3

Overpower AT trips.

When the axial power imbalance is not within the tolerance, the txial power imbalance effect on the Overtemperature and

^

Cverpower AT trips will reduce the Setpoints to provide prctection consistent with core safety limits for cycle specific power distributions.

l LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS (Continued).

Intermediate and Source Rance, Neutron Flux The Intermediate and Source Range, Neutron Flux trips provide core p rotecti or. during reactor startup to mitigate the consequences of an uncon-trolled rod cluster control assembly bank withdrawal from a subcritical condition.

These trips provide redundant protection to the Low Setpoint trip or the Power Range, Neutron Flux channels.

The Source Range channels will initiate a Reactor trip at about 10s counts per second unless manually blocked when P-5 becomes active.

The Intermediate Range channels will initiate a Reactor trip at a current level equivalent to approximately 25% of RATED THERMAL POWER unless manually klocked when P-10 becomes active, overtemoerature ST The Overtemperature aT trip provides core protection to prevent DNB for all comoinations of pressure, power, coolant temperature, and axial power distribu-tion, provided that the transient is slow with respect to pipina transit delays from the core to tne temperature detectors (about 4 seconds), @ pressure is l

within the range between the Pressuri:er High and Low Pressure tripsf The Set-g point is automatically varied with:

(1) coolant temperature to correct for temperature induced changes in density and heat capacity of water and includes dynamic compensation for piping delays from the core to the loop temperature detectors, (2) pressurizer pressure, and (3) axial power distribution. With normal axial power cistribution, this Reactor trip limit is always below the core Safety Limit as shown in Figure 2.1-1. If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the Reactor rio is autom"'cally reduced accordino to the notations in Table 2.2-1.

.Cp we is _Aa ss %n the Ov<.rpoce k[h Overoower aT The Overpower AT trip provides assurance of fuel integrity (e.g., no fuel pellet melting and less than 1% cladding strain) under all possible overpower conditions, limits the required range for Overtemperature aT trip, and provides a backup to the High Neutron Flux trip. The Setpoint is automatically varied with:

(1) coolant temperature to correct for tempera-ture induced changes in density and heat capacity of water, @ (2) rate of l

change of temperature for dynamic compensation for piping delays from the core to the 1000_temoerature detectors,Ito ensure that the allowable heat genera-l ftion rate (kW/ft) is not exceecec.~ The Overpower AT trip provides protection to mitigate the consequences of various size steam breaks as reported in WCAP-0226, " Reactor Core Response to Excessive Secondary Steam Releases."

& cmc { (3) a.yih oW 2-Y bis

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/

h SEABROOK - UNIT 1 B 2-5 qp i

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REACTIVITY CONTROL SYSTEMS BORATION CONTROL i

MODERATOR TEMPERATURE COEpFICIENT LIMITING CONDITION FOR OPERATION 6Ac 3.1.1.3 The moderatoritemperature cesfficient (MTC) shall be within the limits specified in 4.. ;;;; 0^ RATI. C LIMIT 5 ;;?c:1]) COL The maximum N

upper limit shall be less positive thans = t'"g Beginning of cycle life (BOL) limit - MODE 51 and 2* only**.

fPPLICABILITY:

End of cycle life (EOL) limit - MODES 1, 2, and 3 only**.

ACTION:

With the MTC more positive than the BOL limit specified in the COLR, a.

operation in MODES 1 and 2 may proceed provided:

Control rod withdrawal limits are established and maintained 1.

sufficient to restore the MTC to less positive than the BOL t

limit specified in the COLR, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

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

The control rods are maintained within the withdrawal limits 2.

established above until a subsequent calculation veriffes that the MTC has been restored to within its limit for the all rods withdrawn condition; and A Special Report is prepared and submitted to the Commission, 3.

pursuant to Specification 6.8.2, within 10 days, describing the-value of the measured MTC, the interim control red withdrawal limits, and the predicted average core burnup necessary for restoring the positive MTC to within its limit for the all rods withdrawn condition.

With the MTC more negative than the EOL limit specified in the COLR, b.

De in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

(BOL),

+0.5 x 10~' Ak/k/*F for all the rods withdrawn, beginning of cycle life for power levels up to 70% RATED THEFF.AL POWER with a linear ra:::p to 0 Ak/k/'r at 100% RATED THERMAL PCWER.

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

- See Special Test Exceptions Specification 3.10.3.

Amendment No. 9 3/41-4 SEABROOK - UNIT 1 i

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REACTIVITY CCNTROL SYSTEMS

.%VABLE CONTROL ASSEMBLIES R00 ORop TIME i

LIMITING CONDITION FOR OPERATION T

)

3.1.3.4 The individual full-length (shutdown and control) rod drop time from i

the meenanical fully withdrawn positf en snall be less than or equal to G-e seconds from beginning of decay of stationary gripper coil voltage to casnpot i

entry with:

T,yg for each loop greater than or equal to 551*F, and j

a.

b.

All reactor coolant pumps operating.

APPLICABILITY:

MODES 1 and 2.

.j 1'

ACTION-With the drop time of any full-length-rod determined to exceed the above limit, f-3 restore the rod drop time to within the above limit prior to proceeding to

_ DE 1 or 2.

MO SURVEILLANCE REOUIREMENTS 3

4.1.3.4 The red drop time of full-length rods shall be demonstrated through measurement prior to reactor criticality:

i 9

a.

For all rods following each removal of the reactor vessel head, b.

For specifically affected individual rods following any maintenance on cr modification to the Control Rod Drive System that could affect the drop time of those specific rods, and c.

At least once per 18 months.

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1 SEABROOK - UNIT 1 3/4 1-20 j

3/4.2 PCVER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX OIFFERENCE s

g L,ITING CONDITION FOR OPERATION z

the The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained with

3. 2.1 and (flux difference units) about the target flux difference as peci-target fied in e CORE OPERATING LIMITS REPORT (COLR):

ied in The indica. d AFD may deviate outside the required target band spec 4 eater than or equal to 50% but less than 90% of RATE THERMAL the COLR at the indicated AFD is within the Acceptable Operati n Limits POWER provide COLR and the cumulative penalty deviation time ces not exceed specified in th 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months during th previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

The indicated AFD m deviate outside the required target b d specified in the 5% but less than 50% of RATED THERMA POWER provided the COLR at greater than cumulative penalty dev tion time does not exceed 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months uring the previous l

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

APPLICABILITY:

MODE 1, abo IE% of RATED THERMAL P R.*

ACTION:

With the indicated AFD tside of th required target band specified in the COLR and with THE.

L POWER eater than or equal to 90% of a.

!)

RATED THERMAL POWER, withi 15 mi tes either:

f 1.

Restore the indicated AF o within the target band limits, or-l 2.

Reduce THERMAL POWER t les than 90% of PATED THEPtMAL POW b.

With the indicated AFD tside of t, required target band specified in the COLR for more an 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of :. ulative penalty deviation time j

during the previous hours or outsic the Acceptable Operation Limits specified i the COLR and with Th L POWER less than 90% but

(

r than 50% of RATED THE L POWER, reduce:

equal to or grea 1.

THET#.AL R to less than 50% of RATED ERMAL POWER within 30 min es, and nts to less than or ower Range Neutron Flux * ** - High Setp The 2.

next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

eo* 1 to 55% of RATED THERMAL POWER within th "See Specia est Exceptions Specification 3.10.2.

performed ce testing of the Power Range Neutron Flux Channel may ined to Specification 4.3.1.1 provided the indicated AFD is mai

- Surveill A tota of the Acceptable Operation Limits specified in the COLR.

pursua ed ours' operation may be accumulated with the AFD outside of the requ with ion.

target band as specified in the COLR during testing without penalty devi 15 AmendmentNo.9k 3/4 2-1 SEABROOK - UNIT 1

POWER DISTRIBUTION LIMXTS l

3/4.2.1 AXIAL FLUX DIFFERENCE

k.,'i IMITING CONDITION FOR OPERATION

\\

3.2. ACTION

Conti yed) f th the indicated AFD outside of the required target ban as cified in the COLR for more than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of cumulative enalty j

c.

sdev tion time during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and with TF L POWER r

less an 50% but greater than 15% of RATED THERMAL P R, the 1

THERMA POWER shall not be increased equal to or gr er than 50%

of RATE ERMAL POWER.

SURVEILLANCE REQUIRE.'NTS 4.2.1.1 The indicated D shall be determined to be wi in its limits during POWER OPERATION above it - f RATED THERMAL POWER by:

I a.

Monitoring the in ated AFD for each OPE-LE excore channel at least once per 7 da when the AFD Moni Alarm is OPERABLE, and b.

Monitoring and logging he indicated for each OPERABLE excore channel at least once p hour for t first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and at least the AFD Monitor Alarm is once per 30 minutes there "ter, wh he indicated AFD shall be assumed inoperable. The logged va es of to exist during the interval r

ding each logging.

l red outside of its target band when

4. 2.1. 2 The indicated AFD shall be con two or more OPERABLE execre channels e i ' icating the AFD to be outside the target band.

Penalty deviation outs' e of t above required target band shall be accumulated on a time basis of:

One-minute penalty dev' tion for each. minute of POWER OPERATION outside of the targe and at THERMAL P levels equal to or above a.

50% of RATED THERMA ER, and b.

One-half-minute pi alty deviation for each 1 inute of POWER OPERATION outside of the get band at THERMAL POWER 1 els between 15% and 50%'of RATED L POWER.

4.2.1.3 The target f differenca of each OPERABLE execre annel shall be nt at least once per 92 Effective Full ower Days.

determined by measu The provisions-of cification 4.0.4 are not applicable.

t flux difference shall be updated at least once er 4.2.1.4 The ta farence

-Power Days by either determining the target flux 31 Effective F en the cification 4.2.1.3 above or by linear interpolation be le pursuant to S measured value and the predicted value at the end of the most recent revisions of Specification 4.0.4 are not applicable.

life.

Th

/cendment No. O 3/4 2-2 S~

ROOK - UNIT 1 9

3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE LIMITING CONDITION FOR OPERATION j

i 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within:

a.

The limits specified in the COLR. with the Fixed Incore Detector (FIDS) alarm OPERABLE. or b.

The limits specified in the COLR when the FIDS alarm is inoperable.

APPLICABILITY:

MODE 1 above 50% RATED THERMAL POWER.

ACTION:

a.

With the indicated AFD* outside of the applicable limits specified in the COLR:

1.

Either restore the indicated AFD to within the COLR specified limits within 15 minutes, or j

2.

Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux -

High Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , and 3.

THERMAL POWER shall not be increased above 50% of RATED l

THERMAL POWER unless the indicated AFD is within the limits specified in the COLR.

l b.

With an OPERABLE FIDS Alarm exceeding in limit:

l 1.

Comaly with the AFD limits specified in the COLR for operation wit 1 the FIDS Alarm inoperable within 15 minutes and.

2.

Verify THERMAL POWER is less than the maximum power limit established by Surveillance Requirement 4.2.1.2 within 15 minutes and.

3.

Identify and correct the cause of the FIDS alarm prior to i

operation beyond the limits specified in the COLR for operation with the FIDS Alarm inoperable.

c.

With the FIDS Alarm inoperable, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

1.

Comply with the AFD limits specified in the COLR for operation with the FIDS inoperable, and i

2.

Verify THERMAL POWER is less than the maximum power limit establi.shed by Surveillance Requirement 4.2.1.2.

"The indicated AFD shall be considered outside of its limits when two or more OPERABLE excore channels are indicating the AFD to be outside the limits.

SEABROOK - UNIT 1

-3/4 2-1 1

..,.v.

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r---yw,c-v----%

-ww -v v

e-e w-Tv-w-

.--wg wi-o w

--Tq+w gy vwr-m='=y-gr-y

-'e t 't-* 7 m-g et-m

~. _.

i i

3/4.2 POWER DISTRIBUTICN LIMITS i

3/4.2.1 AXIAL FLUX DIFFERENCE SURVEILLANCE REQUIREMENTS j

i 4.2.1.1 The indicated AFD shall be determined to be within its'11mits during POWER OPERATION above 50% of RATED THERMAL POWER by:

a.

Monitoring the indicated AFD for each CPERABLE excore channel at least once per 7 days when the AFD Monitor Alarm is OPERABLE; and, j

b.

Monitoring and logging the indicated AFD for each OPERABLE excore channel at least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and at least

~

once per 30 minutes thereafter, when the AFD Monitor Alarm is inoperable. The logged values of the indicated AFD shall be assumed to exist during the interval preceding each logging.

4.2.1.2 At least once per 31 EFPD, determine the maximum allowed poter for operation with the FIDS Alarm inoperable by comparing F (Z) to the a

F,(Z) limit specified for operation with the FIDS Alarm inoperable.

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a SEABROCK - UNIT 1 3/4 2-2 4

e

POWER DTSTRlBUTTON LlMITS 3/4.2.2 HEATFLUXHOTCHANNELFACTOR-Fg LIMITING CONDITION FOR OPERATION 3.2.2 F (Z) shall be limited by the following relationships:

q F (Z) i F q

K(Z) for P > 0.5 P

F (Z) $ f K(Z) for P < 0.5 q

.5 THERMAL POWER P = RATED THERMAL POWER, and Where:

RTP F

limit at RATED THERMAL POWER (RTP) specified in q = the Fq theCOROPERATINOLIMITSREPORfCOLR)*and K(Z) = the normalized F (Z) as a function of core height q

as specified in the COLR.

APPLICABILITY: MODE 1.

ACTION:

With F (Z) exceeding its limit:

g a.

Reduce THERMAL POWER at least 1% for each 1% F (Z) exceeds the q

limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hcurs; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; subsequent POWER OPERATION may proceed provided the Overpower AT Trip Set-points have been n-educed at least 1% for each 1% F (Z) exceeds the limit, and q

b.

Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced limit re-quired by ACTION a., above; THERMAL POWER may then be increased, provided F (Z) is demonstrated through incore mapping to oc q

within its limit.

SEABROOK - UNIT 1 3/4 2-4 Amendment No. 9

)

POWER DTSTRIBUTION L?MITS 1

l HEAT FLUX HOT CHANNEL FACTOR - F,(Z)

I N

-. 2

~

5-VEILLANCE REOUIREW s

l

4. 2. & ' The provisions of Specification 4.0.411re not applicable.

shall be evaluated r.o determine if F (.Z) is within its mit by:

4.2.2.2 Q

xy ing the movable incore detectors to obtain a power di ribution a.

at any THERMAL POWER greater than E% of RATED THE' L POWER, ma t

b.

Incre ing the measured F component of the power stribution map i

xy by 3% to ccount for manufacturing tolerances and urther increasing the value - E% to account for measurement uncer inti es,

c.

Comparing th F computed (F ) obtained in pecification 4.2.2.2b.,

xy above, to:

1)

The F,y lims s for RATED THERMAL P R (F

) for the appropriate measured core lanes given in Sp ification 4.2.2.2e. and f.,

below, and 2)

The relationship:

L RTP i

p p

gy.

y(

p)),

XY XV 13 is the limi r fractional THERMAL POWER ' operation b

Where F*Y RTP expressed as a fun ion o F

, PF is the Power Factor xy xy Multiplier for F specifie in the COLR and P is the y

fraction of ED TriERMAL POW. at which F was measured.

l xy d.

Remeasuring F according to the folle ing schedule:

1)

When F is greater than the F li it for the appropriate y

mea red core plane but less than the relationship, additional er distribution maps shall be taken a F

compared to F p

x dF either:

XY a)

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding by 20% o RATED THERMAL POWER or greater, the THERMAL POWER at whi F*Cwas last Y

i determined, or b)

At least once per 31 Effective Full-Power Days EFPD),

whichever occurs first.

Amendment No. 9 SEABROOK - UNIT 1 3/4 2-5 e

... ~. - -

y-

- ~

w

\\POVERDISTRIBUTIONLIMITS

~]

T FLUX HOT CHANNEL FACTOR - F.(Z) t SURVE

' NCE REOUIREMENTS s

4.2.2.2d.

Continued)

C 2)

When the F is less than or equal to the F limit f the' x

propriate measured core plane, additional power di ribution ma shall be taken and F compared to F and F ' at least S

x xy once er 31 EFPD.

e.

The F limi. for RATED THERMAL POWER (FRU ) shall e provided for xy xy i

all core planes ontaining Bank "D" control rods a all unrodded core planes in t.

CORE OPERATING LIMITS REPORT p Specification 6.8.1.6, I

The F,y limits of Spe 'fication 4.2.2.2e., ab e, are not applicable f.

in the following core p nes regions as mea red in percent of core height frem the bottcm o the fuel:

l 1)

Lower core region from to 15%, in usive, i;

2)

Upper core region from 85 o 10

, inclusive, i

3)

Grid plane regions at 2*L 8 t

, 32.1 2%, 46.4 ! 2%, 60.6 2%,

and 74.912%, inclusis e, a i

4)

Core plane regions withi

! 2% of ore height (t 2.88 inches) i aeout the bank demand p ition of t.

Bank "0" control rods.

[

g.

With F exceeding F,, t effects of F F (Z) shall be evaluated C

q to determine if F (Z) i w. tc.i n

._ limits.

q 4.2.2.3 When F (Z) is measure for w than F determ tions, an overall q

xy measured F (Z) shall be obt ed from a power distribution m and increased q

by 2% to account for manu cturing tolerances and further inct sed by 5% to i

account for measurement neertainty.

i f

F l

Amendment No. 9 SEA 8R00K - UNIT 1 3/4 2-7

mm mmmnemonour,anmm HEAT FLUX HOT CHANI'EL FACTOR - FfZ1 LIMITING CONDITION FOR OPERATION 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

4.2.2.2 F (Z) shall be demonstrated to be within its 11:

priorto o

operation above 75% RATED THERMAL POWER after ear

'uel loading and at least once per 31 EFPD thereafter by:

Using the incore detector system to obtain a power distribution map a.

at any THERMAL POWER greater than 5% of RATED THERMAL POWER.

b.

Increasing the measured F (Z) component of the power distribution o

map by 3% to account for manufacturing tolerances and further increasing the value by 5% when using the movable incore detectors or 5.21% when using the fixed incore detectors, to account for -

measurement uncertainties.

4.2.2.3 The limits of Specification 3.2.2 are not applicable in the following core plane regions as measured in percent of core height from the bottom of the fuel:

1)

Lower core region from 0 to 15%, inclusive.

2)

Upper core region from 85 to 100% inclusive.

j 4.2.2.4 Each fixed incore detector alarm setpoint shall be updated at least once per 31 EFPD. The alarm setpoints will be based on the latest available power distribution, so that the alarm setpoint does not i

exceed the F (Z) limit defined in Technical Specification 3.2.2.

o s

4 I

SEABROOK - UNIT 1 3/4 2-6 i

PAGE INTENTIONALLY BLANK i

1 SEABROOK - UNIT 1 3/4 2-7

FCWER DISTRIBUTION LIMITS 3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR e

p:x (5ffr

~

LIMITING CCNDITION FOR OPERATION Me Gh* Rmits _sfeer4

'WCOLR

.or 3.2.3 F

snail be less than F.

1.0 + -

q (1-F ;

q CWhere: [F =

MER."

POWER

, and R :D I'

.s".AL POV N

l'.it at * *ED THEF L PCWE7 RTP), - - - -. _-_.

- - - -. _- g- =

..e Fg sp fied in

..e CORE rERATING IMITS R". ORT (C-R),ar

i'd th')

= the cwer F

.or Multi ier fe d 'E'

.g APPLICASILITY:

MODE 1.

ACTION:

i N

WithFiq exceeding its limit:

Within 2 heurs reduce the THERMAL POWER to the level where the a.

LIMITING CONDITION FOR OPERATION is satisfied.

g Identify and c:rrect the cause of the out-of-limit condition prior b.

to increasing THERMAL PCVER a::cve the li::,it required by ACTION a.,

9 acove; THERMAL PCWER may then be increased, previded Fiq is demonstrated thrcugh incere mapping to be within its limit.

SURVEILLANCE REOUIREMENTS A.2.3.1 The previsions of Specification 4.0.4 are not applicable.

I 4.2.3.2 Fiq shall be demonstrated to be within its limit prior to cperation 4

above 75% RATED THERMAL POWER after each fuel loading and at least once per 31 E.:?D thereafter by:

Sq s b x-Using the seveele incere detectordobtain a pcwer distribution

=ap at any THERMAL PCWER greater than 5% RATED THERMAL POWER.

a.

Using the measured value of F'4 which dcas not include an allewance for measurement uncertainty. "q b.

C2h Mene.M nt NC-3 3/4 2-3 SEAERC0K - UNIT 1 1

l i

POWER DISTRIEUTION LIMIT 3 3/4.2.4 CUADRANT POWER TILT RATIO I

)

LIMITING CONDITION FOR OPERATION

'+ W 3.2.4 The QUADRANT POWER TILT RATIO shall not exceed 1.02.

APPLICABILITY:

MODE 1, above 50". of RATED THERMAL POWER *.

ACTION:

With the QUADRANT POWER TILT RATIO determined to exceed 1.02:

a.

Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months reduce THERMAL POWER at least 3". from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1 and similarly reduce the Pcwer Range Neutron Flux-Hign Trip 5etpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and every 7 days thereafter, verify that F (I)h b.

q naTU4.w3 and F] are within their limits by performing Surveil-lance Requirements 4.2.2.2 and 4.2.3.2.

THERMAL POWER and setpoint reductions shall then be in accordance with the ACTION statements of Specifications 3.2.2 and 3.2.3.

SURVEILLANCE REGUIREVENTS t

'O a.2.4.1 The QUADRANT POWER TILT RATIO shall be determined to be within the limit aoove EO% of RATED THERMAL POWER by:

Calculating the ratio at least once per 7 days when the alarm is a.

OPERABLE, anc b.

Calculating the ratio at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during steady-state operation when the alarm is inoperable.

g, 4.2.4.2 The QUADRANT POWER TILT RATIO shall be determined to be within the limit when above 7E% of RATED THERMAL POWER withpto confirm indicated QUAD ne Power Range channel-inoperable by using the -""- incore detectori POWER TILT RATIO at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by either:

d de&c r Using the four pairs of symmetric thinb': locations or a.

detechse b.

Using the.;;;b': incore ;;tccti:n system to conitor the QUADRANT POWER TILT RATIO subject to the require.ments of Specification 3.3.3.2.

.m

'dfj

See' Special Test Exceptions Specification 3.10.2.

g} [

~./

SEAERCOK - UNIT 1 3/4 2-9

e POWER DISTRIBUTION LIMITS 3/4.2.5 DNB PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB-related parameters shall be maintained within the the following limits:

Reactor Coolant System T,yg, 5,5'94.3*F j

a.

b.

Pressurizer Pressure, >

G.

h ;;ter 0;;lant Sy;te; Il;.;, 1, 002,000 @d-P dnseri "TS. 3.2.N')

l APPLICABILITY: MODE 1.

ACTION:

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

i SURVEILLANCE REQUIREMENTS 4.2.5.1 Each of the parameters shown above shall be verified to be within its limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.2.5.2 The RCS flow rate indicators shall be subjected to CHANNEL CALIBRATION at least once per 18 months.

4.2.5.3 The RCS total flow rate shall be determined by a precision heat balance measurement to be within its limit prior to operation above 95% of RATED THERMAL POWER after each fue.1 loading. The provisigns of Specification 4.0.4 are not applicable for entry into MODE 1.

44. r Eerrnd h.si n F. low.

An a..dow awce 4e 3

mea.suremen+ anterisin+v sk).)

be, mak. when comfa. ring mca.surel 4dlow- -lc> TA etmal,3ed h

,3 F.Jl!.e ur.

Limit not. applicable during either a THERMAL POWER ramp in excess of 5% of RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10%

of RATED THERMAL POWER.

[Th. c.ect,ure.

i

"!nchd;; ; 2. 5 :t

-. Low-use& k the R.ev4seel ThermDc'On

- n M1% m u m

~

e 4-Au Amendment No. 14

~

monvu -unu j.

I

)

Insert "TS.3.2.5" J

1 J

c.

Reactor Coolant System Flow shall be-1) 1 382,800 gpm**; and, 2) 1 392,800 gpm***.

i l

1 1

1

.s i

]

I I

i t,

i ABLE 4.3-1 (Continued)

{

TABLE NOTATIONS i

i

~

0nly if the Reactor Trip System breakers happen to be closed and the Control 4

Rod Drive System is capable of rod withdrawal.

4

- Below P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.

- - Below P-10 (Low Setpoint Power Range Neutron Flux Intarlock) Setpoint.

)

(1)

If not performed in previous 31 days.

i' (2)-Comparison of calorimetric to excore power indication above 15% of RATED THERMAL POWER.

Adjust excore channel gains consistent with calorimetric power if aosolute difference is greater than 2%.

The provisions of Specification 4.0.4 are not applicable to entry into MODE 2 or 1.

Single @coint cornparison of incore to excore AXIAL FLUX DIFF (3) aoove of RATED THERMAL POWER.

Recalibrate if the absolute l

S0%

difference is greater than or equal to 3%.

The provisions of l

F.i. tk l

4 Soecification a.O.4 are not acolicable for entrv into MODE 2 or r epe ment, m. <rh A.y s6adQ. m es.w ai.f.es.s-l

' 7 (ryreses d (4) Neutron detectors may be excluded " rom CHANNEL CALIBRATIONbce par 31 EMPb.

Initial plateau curves shall be measured for each detector.

Subsequent (5) plateau curves shall be obtained, evaluated and compared to the initial

.For the Intermediate Range and Power Range Neutron Flux channels curves.

ne provisions of Specification 4.0.4 are not acplicable for entry into MODE 2 or 1.

(5)

Incore - Excore Calibration, above 75% of RATED THERMAL POWER. The provisions of Soecification 4.0.4 are not acclicable-for entry into f

MODE 2 or 1. f).w the gep 5 s et This sur e.AiW.nce t.ep.1rernenY,)

44 S7 thc e p er 92. t y p nf (7) garrerJLg.rA Al m.aAh,cacn trairsnall be tested at least every 62 days i TEST BASIS.

i (8)

(Not used)

(9) Surveillance in MODES 3*,

4*, and 5* shall also include verification that permissives P-5 and P-10 are in their required state for existing plant conditions by observation of the permissive annunciator window.

(10) Setpoint verification is not applicabhe.

(11) The TRIP ACTUATING DEVICE OPERATIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip attachments of the Reactor Trip Breakers.

SEABROOK - UNIT 1 3/4 3-12 i/

i M

O O

O TAulE 3.3-4 m9 ENGINEEREO SAFE 1Y EEAltlRES ACTllA110N SYSTEM INSlRilHENIATION TRIP SETPolNIS 8

SENSOR g

TOTAL ERROR q

FUNCTIONAL UNIT All0WANCE (1 A) Z (S) 1 RIP SE1 POINT All0WABIE VAttlE I.

Safety Injection (Reactor Trip, feedwater Isolation, Start Diesel Generators, Phase "A" Isolation, Containment Ventilation Isolation, and Emergency Feedwater, Service Water to Secondary Component Cooling Water Isolation, CBA Emergency Fan / Filter Actuation, 1

and Latching Relay).

ms

[

a.

Manual Initiation N.A.

N.A.

5 b.

Automatic Actuation Logic H.A.

N.A.

c.

Containment Pressure--Ill-1 4.2

.71 1.67 5 4.3 psig 5 5.3 psig N.A.

.A-

/ oo r*erf74'"]

d.

Pressurizer Pressure--Low 0.99 1 865 psig 3 g gisig e.

Steam Line Pressure--Low 13.1 10.71 1.63

> 585 psig

> 568 psig*

g 2.

Containment Spray a.

Manual Initiation N.A.

N.A.

H.A.

Y-b.

Automatic Actuation Logic N.A.

N.A.

and Actuation Relays

.--4 c.

Containment Pressure--lii-3 3.0 0.71 1.67

< 19.0 psig 5 18.7 psig-l W-

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

s i

TAlllE 3. 3-4 (Continued) h ENGINEERED SAFETY FEAltlRES AcluATION SYSTEM INSTRUMENTAll0H TRIP SETPol 8

SENSOR n

TOTAL ERROR

^

ALLOWANCE (TA) Z (5)

TRIP SETPOINT All0WABLE VALUE FUNCTIONAL UNIT _

Q 9.

Loss of Power (Start g

Emergency feedwater) 4.16 kV Bus E5 and E6 N.A.

N.A.

> 2975

> 2908 volts volts with Uith a < 1.315 a.

Loss of Voltage a < l.20 second time second time delay.

delay.

b.

4.16 kV Bus E5'and E6 N.A.

N.A.

.N.A.

> 3933 volts

> 3902 volts Uith a $ 10 Uith a $ 10.96 l

Degraded Voltage second time second time i

m

)

delay.

Coincident with:

See item 1. above for all Safety injection Trip Setpoints and Safety injection

,m Allowable Values.

10.

Engineered Safety features Actuation System Interlocks-Pressurizer Pressure, P-11 H.A.

N.A.

$ 1950 ps10-5 Psl9 a.

.b.

Reactor Trip, P-4'

'N.A.

N.A.

. N.A.

See item 5. above for all Steam Generator Water Level Trip h

Steam Generator Water Level, c.

.Setpoints and Allowable Values.

P.-14 D

2

..... m.

___.__,,,_.m 2.-..-.-.

...-..._,_.,,,.r.

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

INSTRUMENTATION v

MONITORING INSTRUMENTATON

---.ytnq. :. nc.Te.r. TOR / < V 5 T'E M u-"i-'-

T.

.Q 2; 1s

p-LIMITING CONDITION FOR OPERATION

%.t e ct= r-3.2.2.2 The "..n'.

Incore :e eni:... System s, hall be. OPERAELE with:

l ocamo ^ s. c.nd, a.

At least 7E% of the detector thi;b M,

I cccb c.s$

b.

A minimum of two detector th' 512: per core quadrant,.e#4 2.

a.:

.u.__....---...

...,......._=..%..,.......

7 th::: t'#-t'::.

kd.edM APCLICABILITY: When the ": :b': Incore ::::ti:r System is used for:

a.

Recalibration of the Exccre Neutron Flux Detection System, or b.

Monitoring the QUADRANT POWER TILT RATIO, or c.

Measurement of F and 9

d-d +o the. FID S Alarm nP ACTIO 04tu.= c With the ":;;b': Incore C:t::t':r System inocerable, de not use the system for h

the above acplicable monitoring er calibratien functicas. The previsiens cf V

Specificatien 3.0.3 are not applicable.

SURVEILLANCE REQUIREM:NTS

%%es.-

(Plant precedures are used to determine that the M:';;b': Incore :::::t':r System is OPERABLE.)

v v-m

-p N

s An o?IZAZLI incere detec:c Iccatics shall consist of a fuel asse=bly conta4 4 g a fi=ed detec:cr s :ing with a --d-i~.:= cf three OPEZ1.II.I deteccc:s c an OPIZA3LI =cvable incere detec cr capable of =apping the loca icn.

N m....

'5?.=

i q

f

'd SEAERO0K - UNIT 1 3/4 3-40

,C

!\\D-V

EMERGENCY CORE COOLING SYSTEMS z

ECCS SUBSYSTEMS - T GREATER THAN OR E00AL TO 350*E i

avg SURVEILLANCE REOUIREMENTS 4.5.2 (Continued) d.

At least once per 18 months by:

~'

Verifying automatic interlock action of the RHR system from' the

~

f Reactor Coolant System to ensure that with a simulated or -

1) actual Reactor Coolant System pressure signal greater than'or.

. equal to 365 psig, the interlocks prevent the valves' from being opened.

A visual inspection of the containment sump and verifying that -

l the subsystem' suction inlets are not restricted by debris and.

2) that the sump components (trash racks, screens, etc.)'show no evidence of structural distress or abnormal corrosion.

i At least once per 18 months, during shutdown, by:

~

e.

Verifying that each automatic valve in the flow path actuates d

. to its correct position on'(Safety Injection actuatio 1) 2)

Verifying that each of the following pumps start automatically-l upon receipt of a Safety Injection actuation test signal:-

l a)

Centrifugal charging pump,

~

b)

Safety Injection pump, and c)

RHR pump.

By verifying that each of the following pumps develops the' ind differential pressure on recirculation flow when tested pursuant to f.

Specification 4.0.5:

Centrifugal charging pump,12480 psid; 1)

Safety Injection pump, 1 1445 psid; and 2) 3)

RHR pump, l @ psid.

b i

3/4.5-6 SEAER00K - UNIT 1 Amendment No. 3

EMERGENCY CORE CCOLING SYSTEMS ECCS SUBSYSTEMS - T GREATER THAN OR EOUAL TO 350*F 1

h SURVEILLANCE REOUIREMENTS

~

4.5.2 (Continued)

By verifying the correct position of each electrical and/or i

g.

mechanical position stop for the following ECCS throttle valves:

s 1)

Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months following completion of each valve stroking operation or maintenance on the valve when the ECCS subsystems are required to be OPERABLE, and 2)

At least once per 18 months.

Hich Head SI Svstem Intermediate Head SI System Valve Number Valve Number i

SI-V-143 SI-V-80 SI-V-147 SI-V t SI-V-151 SI-V-104 SI-V-155 SI-V-109 SI-V-117 SI-V-121 SI-V-125 SI-V-129

@b h.

By performing a flow balance test, during shutdown, following l

i completion of modifications to the ECCS subsystems that alter the subsystem flow characteristics and verifying that:

1)

For centr 1tugal charging pump lines, witn a single pump running:

a)

The sum of the injection line flow rates, excluding the highest flow rate, is greater than or equal to G5-r gpm, and g

b) he total pump flow rate is less than or equal-to 2)

For Safety Injection pump lines, with a single pump running:

a)

The sum of the injection line flow rates, excl the highest flow rate,'is greater than or equal to

gpm, ll and j

b) he total pump flow rate is less than or equal to j

f the For RHR pump lines, with a single pump running, the su 3) injection line flow rates is greater than or equal to +869

_.(g 42.13 9PA-Amendment No. 21-SEABROOK - UNIT I 3/4 5-7 e

l i

i

[

\\

l 3/a.2 FCWER DISTRIBUTION LIMITS

..y BASES The specifications of this section provide assurance of fuel integrity h

during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) y (1) maintaining the minimum DNBR in the core greater than or equal q nts by:

to G during normal operation and in short-term transients, and (2) limiting e3 the fission gas release, fuel pellet temcerature, and cladding mecnanical q

properties to witnin assumed cesign criteria.

In addition, limiting the peak e'

linear power density during Condition I events provides assurance that the g

initial conditions assumed for the LOCA analyses are met and the ECCS acceptance g

criteria limit of 2200*F is not exceeded.

x The definitions of certain hot channel and peaking factors as,used in

.D.

these specifications are as follows:

m Heat Flux Hot Channel Factor, is defined as the maximum local heat F (2) q flux on the surface of a fuel rod at core elevation Z divided by the O

average fuel red heat flux, allowing for manufacturing tolerances on C

7 fuel pellets and rods; Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of N

F,.q the integral of linear power along the red with the hignest integrated power to the average red power; and

~

y(Z) / Radial P king Fa er, is efined the r o of ak pc r den

.y I

to av ace cowe densit" in the rizont plane t cor eleva* on Z.

,1 3/a. 2.1 AXIAL FLUX DIFFERENCE l

l

.IAL FLUX DT.ERENCE (AFr assure that. e F (2) upper b and

.h limits on q

limit sp fied in th ORE OPERATIN IMITS REPORT R)

I enve' pe of the q ti s the nor.ali:ed axi ' peaking fa or is not ex ded during ei r normal

,d eration in the ev of xenon distribution llowing power.anges.

aa et flux di" erence is d ermined at ec ilibrium xeno onditions.

N The f*

-length s may be p tiened with' the core in ordance witn l

l

-f the' respectiv insertion

,its and she d be inserte ear their normal l

-S-p ition for saady-state peration at gh power lev The value of the difference tained unde

.hese condit4 s divided by the frac + 'n o.arget flu l

f RATED riERMAL POW is the tar flux differ ce at RATED THERMAL PO for tr associate ore burnuo nditions. Ta et flux differences for ther O

THEF el POWER 1 1s are obt ed by multi 'ying the RATED THERMAL D 'ER value ;

b

.he aopro ate fractic THERMAL PC' level.

The periodic dating of l(

.ne targe ux differer value is n ssary to reflect core rnup l

4 conside

. ions.

I l 0

Although it intended t the plant will be o ated with the AFD

.hin b

pecification 3.2.1 ab. the target flux d

erence, m

the target ban required by use the POWER reductions, co

ol rod motion will uced THERMAL POWER vels. This C.

during rapid ant THERP AFD to dev' te outside,f the target band at -ect the xenon redistr' ution sufficiently

change the

'd deviati will not return to Lenvel of ceakir factors which may be reached on a subseou l

~

Amendment No. 9 B 3/4 2-1 l

SEABROOK - UNIT 1

Inser "B.3/4.2.1" The limits en AXIAL FLUX DIFFERENCE (AFD) specified in the CCRE CPERATING LIMITS REPORT (COLR) assure that the design limits on peak local power density and minimum DNBR are not exceeded during normal operation and the ecnsequences of any Non-LOCA event would be within specified acceptance criteria.

Y(FIDS),

For cperation with the Fixed Incore Detector Syctc..

assurance that the F (Z) limit of Specificatien 3.2.2 is not exceeded during either a

normal cperation or in the event of xenon redistribution following power changes is provided by a separate Fixed Incore Detector Alarm through the plant process computer. A FIDS Alarm will be generated when a predetermined number of individual detectors exceed their alarm setpoint. The setpoint for each individual detector is adjusted by the nor'nal 5.21% for system measurement uncertainty and 3% for engineer! y uncertainty. This assures that the consequences of a LOCA would be within,pecified acceptance criteria.

Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm. The computer deter-mines the 1-minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for two or more CPERABLE execre channel:f are outside the limits specified in COLR.

These alarms are active when power is greater than 50% of RATED THEPMAL POWER.

1

POWER DISTRIBUTION LIMITS BASE 5

-D 3/4.2.1 AXIAL FLUX DIFFERENCE (Continued)

GAir THERM; POWER (

.h the AFn ithin the rget ban provided *,e time d

tion the dev' mion is l'.ited.

Ac dingly, a hour pen '.y deviati n mit c'.alative dring the

.evious 24 urs is pr ided for ration o'.ide of th

.arget

nd but wi n estaoli ed limits ' ile at TH L POWER evels bet' en 50%

.d 90% of D ED THERMA OWER.

F THERMAL Pc R levels etween 1c and 5' of RATED ~ :.RMAL POW, deviatior of the AFr outside o' the tar; and ar ess signi~ cant.

Th enalty of hours' ac* al time r ects th4 reduc signific=

e.

i Provisi-s for moni' ring the A" on an aut atic basi are deri-d from

..e plant ocess camp er througr ne AFD Mon # or Alarm.

The como er deter mines th e minute av rage of e=, of the OPC eBLE exc

. detecto outputs

.a provid an alarm ssage imm

.ately if t AFD for no or mor OPERABL:

exco. channels e outside oe target b d and th (HERM L pc tR is gr ster th 90% of RA*.

THERMAL EWER.

Duri operati at THEsM" POWER 1 sels tween 50% od 90% and etween 15%

.d 50% RA' ' THERMAL WER, tr comout outouts ar slarm mess e wnen the enalty de-ation acc.ulates yond th

[limitsr 1 hcur ar 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , re sectivelyg 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and NUCLEAR ENTHALPY RISE MOT CHANNEL FACTOR The limits on heat flux hot cnannel factor and nuclear enthalpy rise hot channel f actor ensure tnat:

(1) the design limits on peak local power censity and minimum DNBR are not exceeced and (2) in the event of a LOCA, the peak fuel clac temoerature will not exceed the 2200 F ECCS acceptance criteria limit.

Each of these is measuracle but will normally only be determined periccically as specified in Specifications 4.2.2-and 4.2.3.

This periodic surveillance is sufficient to ensure that the limits are maintained provicec:

Control rods in a single group move together with no individual rod a.

insertion differing by more than : 12 steps, indicated, from the group demand position; Control rod groups are sequenced with overlapping groucs as described b.

in Specification 3.1.3.5; The control rod insertion limits of Specifications 3.1.3.5 and c.

3.1.3.5 are maintained; and The axial power distribution, expressed in terms of AXIAL FLUX d.

DIFFERENCE, is maintained within the limits.

[f SEABROOK - UNIT 1 B 3/4 2-2

i

. C'wER OISTE.5UTION L:.w!Ti

~

l 5ASE5 2

. N A.2.2 and 3/4.2.3 HEAT FLUX HOT C'#NNEL FACTOR and NUCLEAR ENTHAL8Y RICE' MOT

' C.-ANNE! FACICR (Continuec)

NFh will be maintained within its limitsgrevided Conditiens a. thr:e: 5 -

d. ateve are maincained.fihe rel tien of F" as a unctier.f Tyg.AL.:C' R

{a11cws[an-s in tne dial pc r sha;e fc all ;. issib

. red i artie limit.

red bewi, reduce =.he value CHER.

Credi* is ava' able t offse Fu

,he se r1e - sins,, *.aling.n 0 i

.1: r uction 1 the gen c =arg1n.

1.aly offe : any re acw penal es. Th' marc'. inclu s :ne f 11cwi s

I-a.

De.gn limi DNER cf 1.:0 vs. 1 3,

hnsA)I"i3,04*2*3 b.

rid s:a ng (X,) c 0.046 v 0.0**

c.

Ther-diffusier c:effici t of 038 vs.

.055, d.

UNE multipli of 0.S6

s. O.

and e.

~ tch reduc en.

7,e a:01 cable val s of ec.cw per ties a refers ed in

.e Fe O

^

. /Onea.suremenij i

Mq_

When an F measurement is taken, an allewanca fer teth -- "---uperrerU q

and sanufacturing tolerance sust be a A nete< g; y ;er-- 4 +

~

n 27 -.qq;;;,7.;

2,- _e.

7 m 'v a full-c:rs sac taken with th In:.

cic:

N _._

> j:

2". allewanca is a:precriata for sanuf acturing ::leranca.

c-u us. -

J w"J

..e clai taK1n.ac cr,,

u.), a u m.

...v,q a

.y,,

se. an tha' the H Channel acter F ' ),

..s wit.k its 1 t.

T' RTF l' it f-RAT-- triERMAL - Ji (F

,, p7,y. ed in CORE ERATT

~

xy y

i 'g IF 5 REF T per Specifi ion 6.8.',, was de ruine res ex-

.ed 3

i m1 neuve ever t5 full en-cf burn c:nd' iens in %c

.I i

J-N is asasured, no ad

d 'Hiballowancas are necessary p 4cr t:

M lKN When RCS F e

i AM M

nd..,33

'y cc=carisen with the established 11..d/xurasuremen errer of e

has 4

been allowed for in detemination f t.

gn DNER value.

/37 I 3/4.2.a CUADRANT POWER TILT RATIO g,,1ew, s p > wev eih-ces v-s t

.ss c.,anges in cere pcwer 8

The purpose of this specf!Teatfen is te detec*' ding normal cperation tne 4

D' sdistribution between acnthlyQnc: ret *'-QUADRANT 'rCan TILT RATI l

limit of incorp

_ speaking fae ces has been established by review ofC ~S.02 is esta i

. - encugn to warrant further investigation.

surved. !c.wcee Aaencnent Mc. !* U SEAERC0K - UNIT 1 3 3/4 2-3

=,1. 3 d c

e

-e-,

-r-e

,m..

~

INSE3.* A

=0vable incere detectors, while 5.2s.': is appropriate for surveillance resu1*:

deter ined with the fixed ine: e detectors.

=. -

i e-i 9

~

Mt s= N e

~..

.J *..

l m..A M_

aim a t e b l_ s I m a ss e s MnPsF*mwe me

f..- -i* e vh. m e.

JA.m..a.

2.

J.

j 9

..J w h whs e..w..

_%. -. _ - _ v_ o M-

_---,.-a d o

n e w m e n.-

n o%a Z:

=me j

I I

i l

l 4

m I

i 1

1 I

t s

l l

d xX m.

3

's-1 1

1 s

?

i

+

,.n m--

,,r3

l l

i I

I Insert "B.3/4.2.3" The design limit DNER includes margin to offset any rod bow penalty.

Margin is also maintained between the safety analysis limit DNER and the l

i design Izmit ONBR.

This margin is available for plant design flexibility, l

i Insert "B.3/4.2.2" For cperation with the Fixed Incore Detector System (FIDS) Alarm i

l l

CPERASLE, the cycle-dependent normalized axial peaking factor, K (2), specified

.n COLR accounts for axial power shape sensitivity in the LOCA analysis.

Assurance that the F (Z) limit of Specification 3.2.2 is met during both a

normal operation and in the event of xenon redistribution following power changes is provided by the FIDS Alarm through the plant process computer.

4his assures that the consequences of a LOCA would be within specified acceptance criteria.

For operation with the FIDS Alarm inoperable, the cycle-dependent normalized axial peaking factor, K (Z), specified in COLR accounts for possible xenon redist-ibution followine; power changes in addition to axial power shape sens;;;vity in the LOCA anal-(sis. This assures that the consequences of a

(

LOCA would be within specified acceptance criteria.

l 1

i 1

a

POWER DISTRIBUTION LIMITS h

BASES t)

t-3/4.2.5 DNB PARAMETERS 4

1 The limits on the DNS related parameters assure that each of the parameters #

is maintained within the normal steady-state envelope of operation assumed in

_j the transient and accident analyses.

The limits are consistent with the E-iti:1J FSAR assumptions and have been analytically demonstrated adequate toS: -t:O C - " - S J 1. 5 Lovvwou d each analyzed transient. Operating procecures

/ include allowances for me rement and indication uncertainty so that the limits and q psig ssurizer are not exceeded.

of 594.3 F for T,yg the m surement.ror of

.4% for total ow rate s based.on per-l[ forming a. recision eat bal ce and u ng the r sult to rmalize he RCS fl rate in ators.

tentia ouling the fee ater ven ri whic sight no be det med coul ias tb result f m the pr ision h balanc in a non n-serva* ve mann Ther ore, a p alty of 0 $ for u etected uling of he fee ter ven ri is lied.

A fouling ich mi '

bias tr RCS flo ate me urement eater *.an 0.1% c be dete ed by m itoring g d trendi vari-plant rforman paramet s.

If d eted, a ion shal (be taken efore erformir subseq-t precis # n heat bi ance mea rements, i.e., ei r the e quant ed and c ensate or in t RCS fl urishal/i effect the fo ing shall he clea to eli ate the culinc cate m curemen or the ve The 12-hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.

The periodic surveillance of indicated RCS flow is sufficient to detect only flow degradation which could lead to operation outside the specified 1

limit.

Cnser "8.3.2.59 i

F b a cc.e P fan c. e c.Pher.

~

f~

a S S M r e.

c.om p k.l a ri c e w-d s

for -

d SEABROOK - UNIT 1 B 3/4 2-4 Amendment No. g, 12

I i

I Insert "3.3.2.5" l

l RCS flow must be greater than or equal to, 1) the Thermal Design Flow (TOF) with an allowance for measurement uncertainty and, 2) the minimum measured flow used in place of the TDF in the analysis of DNB related events when the Revised Thermal Design Procedure (RTDP) mett.odology is utilized.

1 I

i 1

i i

DESIGN FEATURES '

DESIGN PRESSURE KND TEMPERATURE 5'. 2. 2 The containment building is designed and shall be caintained for a 4

maximum internal pressure of 52.0 psig and a temperature of 296*F.

5.3 REACTOR CORE

_ iEmcoruum QNod FUEL ASSEMBLIES q

5.3.1 The core shall contain 193 fuel assemblies with each fuel assemoly con-taining 254 fuel rods clad with Cir m e-S.

Each fuel rod shall have a nominal active fuel length of 144 inches.

The initial core loading shall have a maximum enrichment of 3.15 weight percent U-235.

Reload fuel shall be similar in phy-sical design to the initial core loading and shall have a maximum enrichment of 5.0 weight percent U-235.

CONTROL RCD ASSEMBLIES 5.3.2 The core shall contain 57 full-length control rod assemblies.

The full-length control rod assemolies shall contain a nominal 142 inches of absorber materi al.

The nominal values of absorber material shall be 80% silver,15% in-cium, and 5% cacmium. All control rods shall be clad with stainless steel tubing.

5.4 REACTOR COOLANT SYSTEM DE5IGN PRES 5URE AND TEMPERATURE

5. a.1 The Reactor Coolant System is designed and shall be maintained:

a.

In acccrdance with the Code recuirements specified in Section 5.2 of the F5AR, with allowance for normal decracation pursuant to the a:ciicaole Surveillance Recuirements, b.

For a pressure of 2485 psig, and c.

For a temperature of 550"F, except for the pressurizer which is 680 F.

VOLUME 5.a.2 The total water and steam volume of the Reactor Coolant System is 12,255 cubic feet at a nominal T,yg of 558.5*F.

l 5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorciogical tower shall be located as shown on Figure 5.1-1.

i SEAER00K - UNIT 1 5-5 Amendment No.5

1 ADMINISTRATIVE CONTROLS ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 6.8.1.4 (Continued) to show conformance with 40 CFR Part 190, " Environmental Radiation Protection Standards for Nuclear Power Operation." Acceptable methods for calculating the dose contribution from liquid and gaseous effluents are given in Regulatory Guide 1.109, Rev. 1, October 1977.

The Annual Radioactive Effluent Release Report shall include a list and description of unplanned releases from the sits to UNRESTRICTED AREAS of radioactive materials in gaseous and liquid effluents made during the reporting period.

~

The Annual Radioactive Effluent Release Report shall include any changes made during the reporting period to the PROCESS CONTROL PROGRAM and the ODCM, i

pursuant to Specifications 6.12 and 6.13, respectively, as well as any major change to Liquid, Gaseous, or' Solid Radwaste Treatment Systems pursuant to Specification 6.14.

It shall also include a listing of new locations for dose calculations and/or environmental monitoring identified by the Land Use Census pursuant to Specification 3.12.2.

The Annual Radioactive Effluent Release Report shall also include the following: an explanation as to why the inoperability of liquid or gaseous i

effluent monitoring instrumentation was not corrected within the time specified i

in Specification 3.3.3.9 or 3.3.3.10, respectively; and description of the events leading to liquid holdup tanks or gas storage tanks exceeding the limits of Specification 3.11.1.4 or 3.11.2.5, respectively.

MONTHLY OPERATING REPORTS

\\

6.8.1.5 Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the U.S. Nuclear Regulatory Commission, l

Washington, D.C. 20555, Attn: Document Control Desk, with a copy to the NRC Regional Administrator, no later than the 15th of each month following the calendar month covered by the report.

CORE OPERATING LIMITS REPORT

~

n 6.8.1.6.a Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT prior to each reload cycle, or prior to any remaining portion of a reload cycle, for the following:

i h @

SHUTDOWN MARGIN limit for MODES 1, 2, 3, and 4 for Specification 3.1.1.1, j

h@

SHUTDOWN MARGIN limit for MODE 5 for Specification 3.1.1.2, Moderator Temperature Coefficient BOL and EOL limits, and 300 ppm surveillance limit for Specification 3.1.1.3, 1hS erf "T S. 6. B.I. 6.4

^

SEABROOK - UNIT 1 6-18 Amendment No. N, 22,

1 l

1 Insert "TS.6.8.1.6.a" Cycle dependent overpower AT and overtemperature AT trip setpoint 1.

parameters and function modifiers for operation with skewed axial power profiles for Table 2.2-1 of Specification 2.2.1, 4

1 1

i

+

ADMINISTRATIVE CONTROLS

&c, 6.8.1.6.a (Continued)

Shutdown Rod Insertion limit for Specification 3.1.3.5, Q@

Control Rod Insertion limits for Specification 3.1.3.6, AXIAL FLUX DIFFERENCE limits Ftar;-t 5 5]for Specification lux Hot Channel Factor, F a d th;.:;;;r c::mg a

Qgiti;;1ier 'a for Specification 3 h@

Nuclear Enthalpy Rise Hot Channel Factor, F f"-.

~,.. Q M

......m

.1.'

for Specification 3.2.3.

The CORE OPERATING LIMITS REPORT shall be maintained available in Room.

The analytical methods used to determine the core operating limits 6.8.1.6.b shall be those previously reviewed and approved by the NRC in:

1.

' CAP 272-P

'We ingh se Rei d Saf y Ev #uatic

.eth olo Ju' 1985 W Pro.ieta

,ethod ogy f Spe ficati ns:

L HUTD0' MARG' limi for OES 1 c, 3 d4 l

3.1. '

SHUT WN MA IN li t for DE

3.1

.2 1

.1.3 Mo rator emper are C ffici t

l S tdown.ed In rtio imit j'

.1.3.5 ontro ed Ir artio Limi Nucle Enth py R e Hot hann Facto j

3.1. 3 3.2.

2.

' AP-li.6-P-

" Qual' icati n of a Ph nix-P/ CN' ear si yste for P ssuri d Wat Rea.or Co s" Ju 19 (W Pa r4 ta l

l;i M

odolo for 5 cifi tie :

?

~.1.1.1

- SHUT WH GIN ' mit f MODES

,2 3 an 4

. 1.1.

- SH OWN RGI imit r MOD 3.1.

3

-M erat Te ratur coeffi ent

(

3.

W P-8385

-A, " ower istri tion C tro and ad llo.ng %of -)

t ures T ical por+," Sept er 19

(

repr ~ ta

.i i{f Meth dology.or eciff ions:

hur own Ro Insert' n it 3.

3.5

.1.3.6 C

.rol R Inser on mits IAL F DIFF NC

(

3.2.1 WC 78

,"P Distr' utic Contr s

ho P

u W er eactor,' Dece er 1 (W p spri ry Amendment No. 9 6-18A SEA 8R00K - UNIT 1

l ADMINISTRATIVE CONTROLS 6.8.1.6.b.

(Continued) cations

.ethod gy fo Speci od Inser ion Lir[t 3.1.3. - SF down ntrol od Inse ion Li

,ts 3.1

.6 5.

tter,

.M. And son to. Knei~ (Chief of C e Per ormar e B.-anch, l

RC), ' nuary ', 1980 Attach, nt:

Operati and afet Analysis i

Aspec* of an mprove Lead F low Package Met dolo for Sp ificat n:

XIAL F X DIFF' ENCE 3

.1 6.

.UREG-0, St dard view Plan, U.uclea Reg atory Commis on, Secti 4.3, uclear esign, July 81, Br.ch chnical Posi 1on CPB

.3-1, stingh se Constant xial Of set ntrol (CAOC', Rev.

2, uly 19

_thoda ogy for pecificatio

. 2.1 - AXIA LUX DIFFERE E i

I 7 WC" -7308-L " Evaluation f Nuc1 r Hot hannel Fac r U attain es" December 971 (

repri tary)

Methodo gy for Spec'.icatio.

l 3.2.2 - Heat Flux t Cha el Fac' r

[8.

WC -3622, " West' ghouse CCS Ev uation Med

, Octote, 1975

' rsion," Novem r 197* (W Prop ietary)

Methodology r Spe featic 3.2.2 - He Flux et Chan 1 Factor 9.

WCAP-92 "Wes nghouse CCS Evaluat' n Model February ' 78 lf Versio," Febr-ry 1978 W Propriet

)

Met dology r Speci ication:

3

.2

-H t Flux t Channel F tor prieta[)

10.

4 CAP-791 -P-A, "P wer Peaking actors," January 19 (W P

[

Methe logy fo Specificat# n:

3.2

- Nuc at Enthalpy ise Hot annel Fac rj h @ YAEC-1363-A, "CASMO-3G Validation," April 1988.

YAEC-1659-A, " SIMULATE-3 Validation and Verification," September 1988.

i Amendment No. 9 6-18B SEABROOK - UNIT 1

'" ' ~ ' "

ADMINISTRATIVE CONTROLS

~}

6.8.1.6.b.

(Continued)

Methodology for Specifications:

SHUTDOWN MARGIN for MODES 1, 2, 3, and 4 3.1.1.1 SHUTDCWN MARGIN for MODE 5 3.1.1.2 Moderator Temperature Coefficient 3.1.1.3 Shutdown Rod Insertion Limit

)

3.1.3.5 Control Rod Insertion Limits 3.1.3.6 AXIAL FLUX DIFFERENCE 3.2.1 Heat Flux Hot Channel Factor 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor 3.2.3 f

h Q Seabrook Station Updated Final Safety Analysis Report, Section 15.4.6,

" Chemical and Volume Control System Malfunction That Results in a Decrease in the Baron Concentration in the Reactor Coolant System."

Methodology for Specifications:

SHUTDOWN MARGIN for MODES 1, 2, 3, and 4 3.1.1.1 SHUTDOWN MARGIN for MODE 5 I

3.1.1.2 The core operating limits shall be determined so that all applicable 6.8.1.6.c.

limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as SHUTDOWN MARGIN, and transient and accident The CORE OPERATING LIMITS analysis limits) of the safety analysis are met.

REPORT for each reload cycle, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, to the NRC Document Control Desk with copies to the Regional Administrator and the Resident Inspector.

v 1hsePI IS.6.8.],fo.

l Amendment No. 9 6-18C SEABROOK - UNIT 1

i I

l Insert "TS.6.8.1.6.b" 1.

WCAP-10266-P-A, Rev. 2 with Addenda (Proprietary) and WCAP-11524-A (Nonpr oprietary), "The 1981 version of the Westinghouse ECCS Evaluation Model Using the BASH Code", August, 1986 i

Methodology for Specification:

3.2.2

- Heat Flux Hot Channel Factor 2.

WCAP-10079-P-A (Proprietary) and WCAP-10080-A (Nonproprietary),

"NOTRUMP: A Nodal Transient Small Break and General Network Code",

f August, 1985 Methodology for Specification:

3.2.2

- Heat Flux Hot Channel Factor i

5.

YAEC-1241, " Thermal-Hydraulic Analysis of PWR Fuel Elements Using the CHIC-KIN Code",

R. E. Helfrich, March 1981 l

Methodology for Specifications:

3.2.1

- AXIAL FLUX DIFFERENCE 3.2.2

- Heat Flux Hot Channel Factor 3.2.3

- Nuclear Enthalpy Rise Hot Channel Factor t

6.

YAEC~1849P, " Thermal-Hydraulic Analysis Methodology Using VIPRE-01 For PWR Applications," October ~992 Methodology for Specifications:

2.2.'

'4 4~4ng Safety System Settings 3.2.1

- AXIAL FLUX DIFFERENCE 3.2.2

- Heat Flux Hot Channel Factor t

l 3.2.3

- Nuclear Enthalpy Rise Hot Channel Factor 1

7.

YAEC-1854P, " Core Thermal Limit Protection Function Setpoint Methodology For Seabrook Station," October 1992 Methodology for Specifications:

2.2.1

- Liriting Safety System Settings 3.1.3.5 - Shutdown Rod Insertien Limit 3.1.3.6 - Control Rod Insertion Limits 3.2.1

- AXIAL FLUX DIFFERENCE 3.2.2

- Heat Flux Hot Channel Factor I

3.2.3

- Nuclear Enthalpy Rise Hot Channel Factor 1

I 1

l i

l

?

i Insert "TS.6.8.1.6.b" Cont'd 8.

YAEC-1856P, " System Transient Analysis Methodology Using RETRAN Tor PWR Applications," December 1992 Methodology for Specifications:

2.2.1

- Liniting Safety System Settings 3.1.1.3 - Moderator Temperature Coefficient 3.1.3.5 - Shutdown Rod Insertion Limit 3.1.3.6 - Control Rod Insertion Limits 3.2.1

- AXIAL FLUX DIFFERENCE 3.2.2

- Heat Flux Hot Channel Factor 3.2.3

- Nuclear Enthalpy Rise Hot Channel Factor l

l 9.

YAEC-1752, " STAR Methodology Application for PWRs, Control Rod Ejection, Main Steam Line Break," October 1990 Methodology for Specificatient:

3.1.1.3 - Moderator Tenperature Coefficient 3.1.3.5 - Shutdown Red Insertion Limit 3.1.3. 6 - Control Rod Insertion Limits 3.2.1

- AXIAL FLUX DIFFERENCE 3.2.2

- Heat Flux Hot Channel Factor i

3.2.3

- Nuclear Enthalpy Rise Hot Channel Factor 10.

YAEC-1855P, "Seabrook Station Unit 1 Fixed Incore Detector System Analysis," Octoher 1992 i

Methodology for Specifications:

3.2.1

- AXIAL FLUX DIFFERENCE 3.2.2

- Heat Flux Hot Channel Factor 3.2.3

- Nuclear Enthsipy Rise Hot Channel Factor 11.

YAEC-1642P, " Maine Yankee RPS Setpoint Methodology Using Statistical Ccmbination of Uncertainties - Volume 1 - Prevention of Fuel Centerline Melt," March 1988 Methodology for Specifications:

3.2.1

- AXIAL FLUX DIFFERENCE 3.2.2

- Heat Flux Hot Channel Factor j

3.2.3

- Nuclear Enthalpy Rise Hot Channel Factor l

~1 1

l l

t I

\\

}

Ill.

Retroe of Proposed Chances See attached retype of proposed changes to Technical Specifications. The attached retype reflects the currently issued version of Technical Specifications and also includes the changes proposed in License Amendment Request 92-14, Incore Detector System. Other pending Technical Specification changes or Technical Specification changes issued subsequent to this submittal are not reflected in the enclosed retype. The enclosed retype should be checked for continuity with Technical Specifications pritr to issuance.

Revision bars are provided in the right hand margin to designate a change in the text.

I

~

DEFINITIONS CONTAINMENT INTEGRITY i

1.7 CONTAINMENT INTEGRITY shall exist when:

a.

All penetrations required to be closed'during accident conditions-l are either:

1)

Capable of being closed by an OPERABLE containment automatic isolation valve system, or 2)

Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions.

l 1

b.

All equipment hatches are closed and sealed, l

c.

Each air lock is in compliance with the requirements of' Specification 3.6.1.3.

i d.

The containment leakage rates are within the limits of Specification 3.6.1.2. and e.

The seal-ing mechanism associated with~ each penetration (e'.g..

f welds bellows, or 0-rings) is OPERABLE.

q CONTROLLED LEAKAGE l

1.8 CONTROLLED LEAKAGE shall be that seal water flow supplied to the' reactor f

coolant pump seals.

4 CORE ALTERATION 1.9 CORE ALTERATION shall be the movement or manipulation of any component-l within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATION shall not preclude completion'of movement of a component to a safe conservative position.

1 CORE OPERATING LIMITS REPORT l

1.10 The CORE OPERATING LIMITS REPORT (COLR) provides core operating limits l

4 for the current operating reload cycle. The cycle specific core o)erating limits shall be determined for each reload cycle in accordance wit 1 Specification 6.8.1.6.

Plant operation within these operating limits is addressed in individual specifications.

DIGITAL CHANNEL OPERATIONAL TEST 1.11 A DIGITAL CHANNEL OPERATIONAL TEST shall consist of exercising the

~

digital computer hardware using data base manipulation and/or injecting simulated process data to verify OPERABILITY of' alarm and/or trip functions.

The Digital Channel Operational Test definition is only. applicable to.the Radiation Monitoring Equipment.

i i

SEABR30K - UNIT 1 1-2 Amendment No. 7. 9.

4 l

l c

n 6,IU i

e i

6

.e

.i

.e

.e i

i.

e e

i 8

i i

e i

i I

i e

i i

i 4

*.-----d.-----'----*.----'------"-----'i,-----'e-----"a------'----*.------

660 -

i i

l l

l l UNACC. EPTABLE OPEi.1ATION!

i 650 -------v.

r.-----,-----

r----

r,-----v,-----'------r----

v------

i i

i

.i i.

i i

3 i

s40 -- - - - - - f - - - - - -l- - -

- - - - - - - t. - - - - - i. - - - -. - - - -2;r'po-E- - - i - - - - - -l- - - - - - F - - - - - i. - - - - - -

i S4 t

C i

e i

i i

i

.i t

i i

gy i.

i e

i i.

C a

i i

i 4

.' 225g - - - - '. - - - - - -

3 s 630 --

-- '---- '.----- i i

i i

e i

i i

c i

i i

e i

i e'

i i

i w

i.

e.

i.

i..

i.

by2S pS4.t ----,.

i c.

2

--e-----,i i

r 620 -

r-----v W

--r-i ua i

i i

e i

i c) e.

a i

i i

i "Odo Asq.

i

.i i.

i e

e i

i e

i.

c i

y 610 -- - - - - - i. - - - - -.l - - - - - - l- - - - - - t - - - - - i - - -

--F-----i----

---F-----i-yg60 i

i e

4 i

i a

i i

e i

i i

i.

u) i.

i i

i o

i i

.i c

i 600 ------'.----a-----'----*----d------'-----'.----'.---

'---- A.---

a-i i

e i

i.

i i

i ACCEPTAB.LE OPEF.ATION:

i i

e 590 -- - - - - - +.

r-----e-----m------r-----r-----,.

r-----,i i

i e

i.

i e

i I

i 4

i i

i

- - - - - - i - - - - - -l - - - - - el- - - - - - t - - - - - i - - - - - -l- - - - - - F - - - - - i. - - - - - -l.- - - - - - F. - - -

580 -

i.

i i

i e

i i

i e

i i

e 570 0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 1.1 1.2 FRACTION OF RATED THERMAL POWER FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION SEABROOK - UNIT 1 2-2 Amendment No.

J i

tn TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS n8 SENSOR

^

TOTAL ERROR FUNCTIONAL UNIT ALLOWANCE (TA)

Z (S)

TRIP SETPOINT ALLOWABLE VALUE Z

1.

Manual Reactor Trip N.A.

N.A.

~

2.

Power Range. Neutron Flux a.

High Setpoint 7.5 4.56 0

s109% of RTP*

s111.1% of RTP*

b.

Low Setpoint 8.3 4.56 0

s25% of RTP*

s27.1% of RTP*

3.

Power Range. Neutron Flux.

1.6 0.5 0

s5% of RTP* with s6.3% of RTP* with High Positive Rate a time constant a time constant 22 seconds 22 seconds to 4.

Power Range. Neutron Flux.

1.6 0.5 0

55% of RTP* with 56.3% of RTP* with A.

High Negative Rate a time constant a time constant 22 seconds 22 seconds.

5.

Intermediate Range.

17.0 8.41 0

525% of RTP*

s31.1% of RTP*

Neutron' Flux 5

5 6.

Source Range. Neutron Flux 17.0 10.01

- 0 s10 cps s1.6 x 10 cps 7.

Overtemperature AT N.A.

N.A.

See Note 1 See Note 2 8.

Overpower AT N.A.

N.A.

N.A..See Note 3 See Note 4 9.

Pressurizer Pressure - Low N.A.

N.A.

21945 psig 21.933 psig g

k 10.

Pressurizer Pressure - High' N.A-.

N.A.

N.. A.

s2385 psig 52.397 psig

!?

RTP = RATED THERMAL POWER

?

.m m.

,,m.~

r-

,.m..

.,.~,..-c..

1 TABLE 2.2-1 (continued) m9 REACTOR TRIP SYSTEM INSTRUMENTATION 1 RIP SETPOINTS E8 SENSOR

^

TOTAL ERROR FUNCTIONAL UNIT ALLOWANCE (TA)

Z (S)

TRIP SETPOINT ALLOWABLE VALUE EZ 11.

Pressurizer Water Level - High 8.0 4.20 0.84 s92% of instrument s93.75% of instrument span span 12.

Reactor Coolant Flow - Low 2.5 1.9 0.6 290% of measured 289.3% of measured loop flow loop flow 13.

Steam Generator Water 14.0 12.53 0.55 214.0% of narrow 212.6% of narrow Level Low - Low range instrument range instrument span span 14.

Undervoltage - Reactor 15.0 1 39 0

210.200 volts 29.822 volts Coolant Pumps

?

15.

Underfrequency - Reactor 2.9 0

0

=55.5 Hz 255.3 Hz w

Coolant Pumps

16. Turbine Trip a.

Low Fluid Oil Pressure N.A.

N.A.

2500 psig 2450 psig b.

Turbine Stop Valve N.A.

N.A.

21% open 21% open Closure 17.

Safety Injection Input N.A.

N.A.

_i 3r from ESF e

8-O

.t

..n

,..n..-

, ~. -

~..a v

s,,

,n

~e>

I TABLE 2.2-1 (Continued) v, 92 TABLE NOTATIONS E8 NOTE 1:

OVERTEMPERATURE AT

^

i i

s ATo {Ki - K ((1 + r S)1 + r.S) [T (1 + r S) - T ] + K (P - P ) - f (AI)}

(1) l j

AT- (1 + r S)

(1) 2 3

1 g;

(1 + r 5) (1 + 7 S) s s

2 3

H t

w Measured AT by RTD Instrumentation:

Where:

AT

=

Lead-lag compensator on measured AT:

1+rS

=

i 1+TS2 Time constants utilized in lead-lag compensator for AT. 7 2 8 s.

ri. 72

=

1 r2 s 3 s:

Lag compensator on measured AT:

1 i

1+TS y

3 Time constants utilized in the lag compensator for AT. 73 = 0 s:

T

=

3 I

Indicated AT at. RATED THERMAL POWER:

AT

=

o Value specified in the COLR:

K

=

1 Value specified in the:COLR:

K 2

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

1 + r.S

=

1+rS dynamic compensation; s

?I Time constants utilized in lead-lag compensator for T,y. r,a 33 s.

g T.

r

=

4 s

g r s 4 s:

s

<n A

T Average temperature, F:

=

z-1

= Lag compensato'r on measured T.,,:

t 1*TSs Time constant utilized in the measured T,,, lag compensator r = 0 s:

r

=

s s

l p

TABLE 2.2-1 (Continued) 3; TABLE NOTATIONS 8g NOTE 1:

(Continued)

T s 588.5 F (Nominal T at RATED THEFMAL POWER):

m Value specified in COLR:

K

=

3 Pressurizer pressure. psig:

j P

=

2235 psig (Nominal RCS operating pressure):

P

=

i Laplace transform operator, s'l:

i S

=

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

power-range neutron ion chambers as specified in the COLR.

NOTE 2:

Cycle dependent values for the channel's Allowable Value are specified in the COLR.

7 en

$g 8

<+

2 P

.M 1

I f

.7

-____-, -.ww.e-,wm-.www,ww,-,

wn-,,.w--i,yw a _vww vi we ww -e.ewww,,

-+tw-w,,wr,3%wm,+-w+sr.m-+-,w-ere--,,w. a w -w1 v+,

e e rww w

,.ec.,m4.wm-.uwe.w.,_.--.,,.-e.,,,-.-m,-,.

,.._ caw

[

a TABLE 2.2-1 (Continued) v, 9

TABLE NOTATIONS (Continued) i E8 NOTE 3:

OVERPOWER AT n

(1[r7S) (1 + rTS) T - K

( }

[T (1 + r 5) - T1 - f (AI)}

s ATo {K - K 2

4 5

6 (1

r S) 3 3

Z s

As defined in Note 1.

Where:

AT

=

1 + rd

= As defined in Note 1.

1+TS2

= As defined in Note 1.

T.72 1

As defined in Note 1.

1

=

1 + r,5 r

=- As defined in Note l'.

3 As defined in Note 1.

AT

=

o Value specified in the COLR.

K

=

4 Value specified in the 'COLR.

K

=

s The function generated by.the rate-lag compensator for T.,o 1+75 compensation.

~

dynamic-7S

=

7 7

r7~

Time constants-utilized in rate-lag compensator:for.T.y,. 7 = 10 s.

=

7 y

1

= - As. defined in Note 1.

1+rS s

af As defined in Note 1.

r

=

s

A TABLE 2.2-1 (Continued) g; TABLE NOTATIONS (Continued) x k

NOTE 3:

(Continued)

K Value specified in COLR,

=

3 E

As defined in Note 1.

H T

=

T' 2

Indicated T at RATED THERMAL POWER (Calibration temperature for AT

=

instrumentaly1on, s 588.5 F),

t As defined in Note 1. and S

=

i A function of the indicated-diffe'rence between the top and bottom detectors of the j

f (al) 2 power-range neutron ion chambers as specified in the COLR.

NOTE 4:

Cycle dependent values for the channel's Allowable Value are specified in the COLR.

i m

l

+

a t

f i

8 8-8 r

Z-i C

i 1

4 6,,-i

-e-...--..

---.----i4.-.e,-+w.,-r,.v.,,--emn-

-r..*-

w e.**>=----.mme

--e eem

--r

-

w wm*=-war

,m--

=e-wwon-*.=

.e,----eww v-n**

e.<w-+-w w n-e e - ~ww r-m..nra-.45 &+w,-

, e 4

=w-wwww.aseweo=,%ww.---,--+a.www-..~--

2.1 SAFETY LIMITS s

BASES 2.1.1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation that would result in the release of fission 3roducts to the reactor coolant.

Overheating of the fuel cladding is prevented

]y restricting fuel operation to within the nucleate boiling regime where the j

heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and.

therefore. THERMAL POWER and reactor coolant temperature and pressure have been related to DNB. This relation has been developed to predict the DNB flux and l

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

The local DNB heat flux ratio (LNBR) is defined as the ratio of the heat flux -

+

that would cause DNB at a particular core location to'the local heat flux and is indicative of the margin to DNB.

The DNB design basis is as follows:

uncertainties in the DNBR correlation, plant operating parameters, nuclear and thermal parameters, fuel fabrication parameters and computer codes are considered statistically such that there is at least a 95 3ercent probability with 95 percent confidence level i

that DNB will not occur on t1e most limiting fuel rod during Condition I and II i

events.

This establishes a design DNBR value which must be met in plant safety I

analyses using values of input )arameters without uncertainties.

In addition.

i margin has been maintained in t1e design by meeting safety analysis DNBR limits in performing safety analyses.

The curves of Figure 2.1-1 show the loci of points of THERMAL POWER.

Reactor Coolant System pressure. and average temperature for which the minimum DNBR is no less than the safety analysis DNBR limit value, or the average g

enthalpy at the vessel exit is equal to the enthalpy of saturated liquid.

These curves are based on an enthalpy rise hot channel factor Flg, at WD THERMAL POWER. of 1.65. The value of FL at reduced power is assumed to vary according to the expression:

FL=1.65[1+0.3(1-P)]

Where P is the fraction of RATED THERMAL POWER.

Thisexpressionconservativelyboundsthecycles)ecificlimitsonFL sper.ified in Technical Specification 3/4.2.3 and the C0_R. The Safety Limits in Figure 2.1-1 are also based on a reference cosine axial power shape with a peak i

of 1.55.

t I

)

l SEABROOK - UNIT 1 B 2-1

-Amendment No.

- ~

SAFETY LIMITS BASES i

j i

2.1.1 REACTOR CORE (Continued) i The resulting heat flux conditions are more limiting than those calculated.

for the range of all control rods fully withdrawn to the maximum allowable.

control rod insertion. assuming the axial power imbalance is within the limits of the f (AI) and f (AI) functions of the 0vertemperature and Overpower AT trips.

3 2

4 1

When the axial power imbalance is not within the tolerance.-the axial power imbalance effect on the Overtemperature AT and Overpower AT trips will reduce.

1 the setpoints to provide protection consistent with core safety limits for cycle specific power distribution.

2.1.2 REACTOR COOLANT SYSTEM PRESSURE i

The restriction of this Safety Limit protects the integrity of the Reactor Coolant System (RCS) from overpressurization and thereby prevents the release of i

radionuclides contained in the reactor-coolant from reaching the containment

~ atmosphere ^;

-~

The reactor vessel, pressurizer, and the RCS piping, valves, and fittings dre designed to Section III of the ASME Code for Nuclear Power Plants, which permits a maximum transient pressure of 110% (2735 psig) of design pressure.

The Safety Limit of 2735 psig is, therefore, consistent with the design criteria and associated Code requirements.

i The entire RCS is hydrotested at 125% (3110 psig) of design pressure to demonstrate integrity prior to initial operation.

~

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l SEABROOK UNIT 1 B 2-2 Amendment No.

i LIMITING SAFETY SYSTEM SETTINGS J

BASES l

l 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS (Continued) l Intermediate and Source Ranoe. Neutron Flux i

The Intermediate and Source Range. Neutron Flux trips provide core protection during reactor startup to mitigate the consequences of an uncon-1 trollea roa cluster control assembly bank withdrawal from a subcritical condition.

These trips provide redundant 3rotection to the Low Setpoint trip of the Power Range. Neutron Flux channels.

T1e Source Range channels will initiate 5

a Reactor trip at about 10 counts per second unless manually blocked when P-6 becomes active. The Intermediate Rdnge Channels Will initiate a Reactor trip'at a current level equivalent to approximately 25% of RATED THERMAL POWER unless manually blocked when P-10 becomes active.

Overtemoerature AT i

all combinations of pressure, power; coolant temperature. and axial pow.NB for The Overtemperature AT trip provides core protection to prever' D er j

distribution, provided that the transient is slow with respect to pi delays from the core to the temperature detectors (about 4 seconds)-, ping transit l

pressure is within the range between the Pressurizer High and Low Pressure trips and power is less than the Overpower AT trip setpoint. The Setpoint is automatically varied with: (1) coolant temperature to correct for temperature induced changes j

in density and heat capacity of water and includes dynamic compensation for-piping delays from the core to the loop temperature detectors. (2) pressurizer pressure, and (3) axial power distribution. With normal axial power distribution. this Reactor trip limit is always below the core Safety Limit as shown in Figure 2.1-1.

If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the Reactor trip is automatically reduced according to the notations in Table 2.2-1.

Overpower AT The Overpower AT trip provides assurance of fuel integrity (e.g., no fuel pellet melting and less than 1% cladding strain) under all possible overpower conditions. limits the required range for Overtemperature AT trip, and provides a backup to the High Neutron Flux trip. The Setpoint is automatically varied with: (1) coolant temperature to correct for temperature induced changes in density and heat capacity of water. (2) rate of change of temperature for i

l dynamic compensation for piping delays from the core to the loop temperature detectors, and (3) axial power distribution to ensure that the allowable heat l

generation rate (kW/ft) is not exceeded. The Overpower AT tri) provides protection to mitigate the consequences of various size steam areaks as reported in UCAP-9226. " Reactor Core Response to Excessive Secondary Steam Releases."

SEABROOK - UNIT 1 B 2-5 Amendment No.

REACTIVITY CONTROL SYSTEMS BORATION CONTROL MODERATOR TEMPERATURE COEFFICIENT l

LIMITING CONDITION FOR OPERATION l

3.1.1.3 The moderator temperature coefficient (MTC) shall be within the limits specified in the COLR. The maximum upper limit shall be less positive than +0.5 X 10" Ak/k/ F for all the rods withdrawn. beginning of cycle life (BOL), for power levels up to 70% RATED THERMAL POWER with a linear ramp to 0 Ak/k/ F at 100% RATED THERMAL POWER.

i APPLICABILITY:

Beginning of cycle life (BOL) limit - MODES 1 and'2* only**.

End of cycle life (EOL) limit - MODES 1. 2. and 3 only**

ACTION:

a.

With the MTC more positive than the BOL limit specified in the COLR, operation in MODES 1 and 2 may proceed provided:

1 1.

Control rod withdrawal limits are established and maintained sufficient to restore the MTC to less positive than the BOL limit specified in the COLR, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . These withdrawal limits shall be in addition to the insertion limits of Specification 3.1.3.6:

2.

The control rods are maintained within the withdrawal limits established above until a subsequent calculation verifies that the MTC has been restored to within its limits for the all rods withdrawn condition: and i

3.

A Special Report is prepared and submitted to the Commission.

pursuant to Specification 6.8.2. within 10 days, describing the value of the measured MTC. the interim control rod withdrawal limits. and the predicted average core burnup necessary for restoring the positive MTC to within its limit for the all rods withdrawn condition.

b.

With the MTC more negative than the EOL limit specified in the COLR, be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

With k,ff greater than or equal to 1.

- See Special Test Exceptions Specification 3.10.3.

i SEABROOK - UNIT 1 3/4 1-4 Amendment No. 9,

REACTIVITY CONTROL SYSTEMS j

MOVABLE CONTROL ASSEMBLIES ROD DROP TIME I

LIMITING CONDITION FOR OPERATION 3.1.3.4 The individual full-length (shutdown and control) rod drop time from the mechanical fully withdrawn position shall be less than'or equal to 2.4 li seconds from beginning of decay of stationary gripper' coil ' voltage.to dashpot entry with:

l l

a.

T for each loop greater than or equal to 551 F and j

m b.

All reactor coolant pumps operating.

APPLICABILITY: MODES 1 and 2.

ACTION-l i

With the drop time of any full-length rod determined to exceed the above limit.

]

restore the rod drop time to within the. above limit prior to proceeding to MODE 1 or 2.

l SURVEILLANCE REQUIREMENTS j

4.1.3.4 The rod drop time of full-length rods shall-be demonstrated through i

measurement prior to reactor criticality:

)

e a.

For all rods following each removal of the reactor vessel head.

t b.

For specifically affected individual rods following any maintenance on or modification to the Control Rod Drive System that could affect the drop time of those specific rods, and c.

At least once per 18 months.

f J

SEABROOK - UNIT 1 3/4 1-20 Amendment No. 8.

1

..._m..~~..,,

UI

3/4 2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained-within:

a.

The limits specified in the COLR. with the Fixed Incore Detector (FIDS) Alarm OPERABLE or b.

The limits specified in the COLR. when the FIDS Alarm is inoperable.

APPLICABILITY.

MODE 1 above 50% RATED THERMAL' POWER.

i ACTION:

I

't a.

With the indicated AFD* outside of the applicable limits specified in the COLR:

1.

Either restore the indicated AFD to within the COLR specified i

limits within 15 minutes, or 2.

Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux -

i High Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , and 3.

THERMAL POWER shall not be increased above 50% of RATED THERMAL POWER unless the indicated AFD is within the limits specified in the COLR.

I b.

With an OPERABLE FIDS Alarm exceeding in limit:

1.

Com)ly with the AFD limits specified in the COLR for operation wit 1 the FIDS Alarm inoperable within 15 minutes and.

2.

Verify THERMAL POWER is less than the maximum power limit established by Surveillance Requirement 4.2.1.2 within 15 minutes and.

3.

Identify and correct the cause of the FIDS Alarm prior to i

operation beyond the limits specified in the COLR for operation with the FIDS Alarm inoperable.

c.

With the FIDS Alarm inoperable, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

1.

Comaly with the AFD limits specified in the COLR for operation wit 1 the FIDS Alarm inoperable, and 2.

Verify THERMAL POWER is less than the maximum power limit established by Surveillance Requirement 4.2.1.2.

The indicated AFD shall be considered outside of its limits when two or more OPERABLE excore channels are indicating the AFD to be outside the limits.

SEABROOK - UNIT 1 3/4 2-1 Amendment No. 9.

3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE SURVEILLANCE REQUIREMENTS 4.2.1.1 The indicated AFD shall be determined to be within its limits during i

POWER OPERATION above 50% of RATED THERMAL POWER by:

a.

Monitoring the indicated AFD for each OPERABLE excore channel at least once per 7 days when the AFD Monitor Alarm is OPERABLE, and b.

Monitoring and logging the indicated AFD for each OPERABLE excore channel at least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and at least once per 30 minutes thereafter, when the AFD Monitor Alarm is inoperable. The logged values of the indicated AFD shall be assumed to exist during the interval preceding each logging.

l 4.2.1.2 At least once per 31 EFPD determine the maximum allowed power for operation with the FIDS Alarm inoperable by comparing F (Z) to the i

o F (Z) limit established for operation with the FIDS Alarm oinoperable.

3 t

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SEABRDOK - UNIT 1 3/4 2-2 Amendment No. 9.

)

POWER DISTRIBUTION LIMITS 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - F (Z) e LIMITING CONDITION FOR OPERATION 3.2.2 F (Z) shall be limited by the following relationships:

o Fo(Z) 5 F$" K(Z) for P > 0.5 l

P i

F (Z) 5 F$" K(Z) for P 5 0.5 o

.5 Where:

P=

THERMAL POWER

, and RATED THERMAL POWER l

F$"

=

the F limit at RATED THERMAL POWER (RTP) l specikied in the COLR. and K(Z) the normalized Fo(Z) as a function of core height

=

as specified in the COLR.

l t

APPLICABILITY:

MODE 1.

ACTION:

I l

a.

With F (Z) exceeding its limit-f o

1.

Reduce THERMAL POWER at least 1% for each 1% F (Z) exceeds the o

limit within 15 minutes' and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months s:

POWER OPERATION may proceed for up to a total of 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months s:

subsequent POWER OPERATION may proceed provided the Overpower

{

AT Trip Setpoints have been reduced at least 1% for each 1%

F (Z) exceeds the limit, and j

o 2.

Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced limit required by ACTION a.. above: THERMAL POWER may then be j

increased.

3rovided F (Z) is demonstrated through incore n

mapping to )e within its limit.

l l

SEABROOK - UNIT 1 3/4 2-4 Amendment No. 9.

1 I

..=

POWER DISTRIBUTION LIMITS HEATFLUXHOTCHANNELFACTOR-Ffll 1

l LIMITING CONDITION FOR OPERATION 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

[

t 4.2.2.2 F (Z) shall be demonstrated to be within its limits prior to 4

o operation above 75% RATED THERMAL POWER after each fuel loading and at least once per 31 EFPD thereafter by:

i a.

Using the incore detector system to obtain a power distribution map at any THERMAL POWER greater than 5% of RATED THERMAL POWER.

i i

b.

Increasing the measured Fo(Z) component of the power distribution map by 3% to account for manufacturing tolerances and further increasing the value by 5% when using the movable incore detectors j

or 5.21% when using the fixed incore detectors, to account for measurement uncertainties 4.2.2.3 The limits of Specification 3.2.2 are not applicable in the following core plane regions as measured in percent of core height from the bottom of the fuel:

1)

Lower core region from 0 to 15%, inclusive.

l 2)

Upper core region from 85 to 100%. inclusive.

4.2.2.4 Each fixed incore detector alarm setpoint shall be updated at least once per 31 EFPD. The alarm setpoints will be based on the latest

.i available power distribution, so that the alarm setpoint does not i

exceed the F (Z) limit defined in Technical Specification 3.2.2.

o l

I 1

SEABROOK - UNIT 1 3/4 2-6 Amendment'No. 9.

t i

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1

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3

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+

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t PAGE INTENTIONALLY LEFT BLANK.

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SEABROOK - UNIT 1 3/4 2-7 Amendment No. 9.

i POWER DISTRIBUTION LIMITS l

3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR LIMITING CONDITION FOR OPERATION i

i 3.2.3 Fj shall be less than the limits specified in the COLR.

APPLICABILITY:

MODE 1.

l l

ACTION:

l With 5l exceeding its limit:

l a.

Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months reduce the THERMAL POWER to the level where the LIMITING CONDITION FOR OPERATION is satisfied.

j b.

Identify and correct the cause of the out-of-limit condition prior i

to increasing THERMAL POWER above the limit required b ACTION a..

i above: THERHAL POWER may then be increased, provided

-is i

demonstrated through incore mapping to be within its imi t.

i SURVEILLANCE REOUIREMENTS i

~

4.2.3.1 The provisions of Specification 4.0.4 are not applicable.

i above 75%'O shall be demonstrated to be within'its limit prior to operation 4.2.3.2 F i

kATED THERMAL POWER after each fuel loading anc at least once per 31 EFPD thereafter by.:

a.

Using the incore detector system to obtain a power distribution map at any THERMAL POWER greater than 5% RATED THERMAL POWER.

l b.

Using the measured value of FS which does not include an allowance for measurement uncertainty.

3

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I 7

i SEABROOK - UNIT 1 3/4 2-8 Amendment No. 9.

POWER DISTRIBUTION LIMITS 3/4.2.4 OUADRANT POWER TILT RATIO LIMITING CONDITION FOR OPERATION i

3.2.4 The OUADRANT POWER TILT RATIO shall not exceed.1.02.

I APPLICABILITY: MODE 1. above 50% of RATED THERMAL POWER *.

l ACTION:

l i

With the OUADRANT POWER TILT RATIO determined to exceed 1.02:

l a.

Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months reduce THERMAL POWER at least 3% from RATED THERMAL.

i POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess-of 1 and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next.4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

l b.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and every 7 days thereafter, verify that F (Z) and I

l q

FL are within their limits by aerforming Surveillance Requirements 4.2.2.2 and 4.2.3.2.

THERMAL 30WER and setpoint reductions shall i

then be in accordance with the ACTION statements of Specifications 3.2.2 and 3.2.3.

]

SURVEILLANCE REQUIREMENTS j

l e

i i

i 4.2.4.1 The OUADRANT POWER TILT RATIO shall be determined to be within the i

limit above 50% of RATED THERMAL POWER by-

~t a.

Calculating the ratio at least once per 7 days when the alarm is

'f OPERABLE. and j

b.

Calculating the ratio at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during steady-state i

operation when the alarm is inoperable.

i 4.2.4.2 The OUADRANT POWER TILT RATIO shall be determined to be within the limit when above 75% of RATED THERMAL POWER with one Power Range channel inoperable by using the incore detector system to confirm indicated OUADRANT i

j POWER TILT RATIO at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by either:

a.

Using the four pairs of symmetric detector locations or I

b.

Using the incore detector system to monitor the OUADRANT POWER TILT RATIO subject to the requirements of Specification 3.3.3.2.

i t

i

See Special Test Exceptions Specification 3.10.2 SEABROOK - UNIT 1 3/4 2-9 Amendment No.

POWER DISTRIBUTION LIMITS 3/4.2.5 DNB PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB-related parameters shall be maintained within the following limits:

a.

Reactor Coolant System Tm. s 594.3 F 1

I b.

Pressurizer Pressure. 2 2185 psig*

c.

Reactor Coolant System Flow shall be:

1.

2 382.800 gpm**; ano.

2.

2 392.800 gpm***

APPLICABILITY:

MODE 1.

i ACTION:

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

SURVEILLANCE REQUIREMENTS 4.2.5.1 Each of the parameters shown above shall be verified to De within its limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.2.5.2 The RCS flow rate indicators shall be subjected to CHANNEL CALIBRATION at least once per 18 months.

4.2.5.3 The RCS total flow rate shall be determined by a precision heat balance measurement to be within its limit prior to operation above 95% of RATED THERMAL POWER after each fuel loading. The provisions of Specification 4.0.4 are not applicable for entry into MODE 1.

Limit not applicable during either a THERMAL POWER ramp in excess of 5% of RATED THEPJiAL POWER per minute or a THERMAL POWER step in excess of 10%

of RATED THERMAL POWER.

- Thermal Design Flow. An allowance for measurement uncertainty shall be made when comparing measured flow to Thermal Design Flow.

- - Minimum measured flow used in the Revised Thermal Design Procedure.

SEABROOK - UNIT 1 3/4 2-10 Amendment No. -14.

I TABLE'4.3-1 (Continued) l i

j TABLE NOTATIONS j

i i

0nly if the Reactor Trip System breakers happen to be closed and the Control i

j Rod Drive System is capable of rod withdrawal.

j i

- Below P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.

l i

]

- - Below P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.

(1)

If not performed in previous 31 days.

(2)

Ccmparison of calorimetric to excore power indication above 15% of RATED THERMAL POWER. Adjust excore channel gains consistent with calorimetric power if absolute difference is greater than 2%. The provisions of i

Specification 4.0.4 are not applicable to entry.into MODE 2 or 1.

l (3)

Single point comparison of incore to excore AXIAL FLUX DIFFERENCE above l

50% of RATED THERMAL POWER.

Recalibrate if the absolute difference is l

greater than or equal to 3%. The provision.s of Specification 4.0.4 are j

not applicable for entry into MODE 2 or 1.

For the purposes of this I

surveillance requirement, monthly shall mean at least once per 31 EFPD.

(4)

Neutron detectors may be excluded from CHANNEL CALIBRATION.

-i (5)

Initial plateau curves shall be measured for each detector.

Subsequent plateau curves shall be obtained, evaluated and compared to the initial curves.

For the Intermediate Range and Power Range Neutron Flux channels i

the provisions of Specification 4.0.4 are.not applicable for entry into MODE 2 or 1.

(6)

Incore - Excore Calibration. above 75% of RATED THERMAL POWER. The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

For the purposes of this surveillance requirement, quarterly shall mean at least once per 92 EFPD.

i i

(7)

Each train.shall be tested at least every 62 days on a STAGGERED TEST l

BASIS.

[

l i

(8)

(Not used)

(9)

Surveillance in MODES 3*. 4*, and'5* shall also include verification that permissives P-6 and P-10 are in their required state for existing plant i

conditions by observation of the permissive annunciator window.-

(10) Setpoint verification is not applicable.

l (11) The TRIP ACTUATING DEVICE OPERATIONAL TEST shall independently verify the' I

OPERABILITY of the undervoltage and shunt trip attachments of the Reactor Trip Breakers.

j SEABROOK - UNIT 1 3/4 3-12 Amendment No.

1

p TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS 7

SENSOR TOTAL ERROR c

5 FUNCTIONAL UNIT ALLOWANCE (TA) Z (S)

TRIP SETPOINT ALLOWABLE VALUE H

~ 1.

Safety Injection (Reactor Trip.

Feedwater Isolation. Start Diesel Generator. Phase "A" Isolation.

Containment Ventilation Isolation, and Emergency Feedwater. Service Water to Secondary Component Cooling Water Isolation. CBA Emergency Fan / Filter Actuation, and Latching Relay).

y a.

Manual Initiation N.A.

N.A.

?'

b.

Automatic Actuation Logic N.A.

N.A.

T c.

Containment Pressure--Hi '

4.2 0.71 1.67 s 4.3 psig 5 5.3 psig d.

Pressurizer Pressure--Low N.A.

N.A.

= 1300 psig

= 1786 psig e.

Steam Line Pressure--Low 13.1 10.71 1.63

= 585 psig a 568 psig*

2.

Containment Spray a.

Manual Initiation N.A.

N.A.

b.

Automatic Actuation Logic N A.

N A.

N.A.

k and Actuation Relays a

c.

Containment Pressure--Hi-3 3.0 0.71 1.67 s 18.0 psig s 18.7 psig A

EF

TABLE 3.3-4 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS x

8 SENSOR TOTAL ERROR y

FUNCTIONAL UNIT ALLOWANCE (TA) Z (S)

TRIP SETPOINT ALLOWABLE VALUE 5

H 9.

Loss of Power (Start Emergency Feedwater) a.

4.16 kV Bus E5 and E6 N.A.

N.A.

2 2975 2 2908 volts Loss of Voltage volts with with a s 1.315 a s 1.20 second time second time delay.

delay.

b.

4.16 kV Bus E5 and E6 N.A.

N.A.

2 3933 volts a 3902 volts Degraded Voltage with a 5 10 with a s 10.96 W

second time second time delay.

delay.

A ca Coincident with:

Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and Allowable Values.

10. Engineered Safety Features Actuation System Interlocks a.

Pressurizer Pressure. P-11 N.A.

N.A.

s 1950 psig s 1962 psig l

b.

Reactor Trip. P-4 N.A.

N.A.

c.

Steam Generator Water Level.

See. Item 5. above for all Steam Generator Water Level-' Trip

,E P-14 Setpoints and Allowable Values.

I a

E

INSTRUMENTATION i

MONITORING INSTRUMENTATION l

INCORE DETECTOR SYSTEM LIMITING CONDITION FOR OPERATION 3.3.3.2 The Incore Detector System shall be OPERABLE with:

l i

a.

At least 75% of the detector locations and, b.

A minimum of two detector locations per core quadrant, I

c.

An OPERABLE incore detector location consist of a fuel assembly containing a fixed detector. string with a minimum of three OPERABLE-detectors or an OPERABLE movable incore detector capable of mapping the location.

APPLICABILITY:

When the Incore Detector System.is used for:

a.

Recalibration of the Excore Neutron Flux Detection System. or b.

Monitoring the OUADRANT POWER TILT RATIO. or -

c.

Measurement of FL and F (Z). or o

d.

Input into the FIDS Alarm i

ACTION:

With the Incore Detector System inoperable, do not use the system for the above applicable monitoring or calibration functions. The provisions of Specification 3.0.3 are not applicable SURVEllLANCE REQUIRE;4ENTS (Plant procedures are used to determine that the Incore Detector System is OPERABLE.)

SEABROOK - UNIT 1 3/4_3-40 Amendment No.

t EMERGENCY CORE COOLING SYSTEMS i

ECCS SUBSYSTEMS - T GREATER THAN CR EOUAL TO 350 F x

SURVE2 LANCE REQUIREMENTS l

4.5.2 (Continued) l I

d.

At least once per 18 months by:

j 1)

Verifying automatic interlock action of the RHR system from the Reactor Coolant System to ensure that with a simulated or actual Reacter Coolant System pressure signal greater than or i

equal to 365 psig, the interlocks prevent the valves from j

being opened.

2)

A visual inspection of the containment sump and verifying that the subsystem suction inlets are not restricted by debris and I

that the sump components (trash racks, screens, etc.) show no evidence of structural distress or abnormal corrosion.

l e.

At least once per 18 months, during shutdown by; 1)

Verifying that each automatic valve in the-flow path actuates i

to its correct position on (Safety Injection actuation and i

Automatic Switchover to Containment Sump) test signals, and l

2)

Verifying that each of the following pumps start automatically upon receipt of a Safety Injectics actuation test signal:

l a)

Centrifugal charging pump, I

i j

b)

Safety Injection pump. and l

c)

RHR pump.

l 1

f.

By verifying that each of the following pumps develops the indicated i

differential pressure on recirculation flow when tested pursuant to i

Specification 4.0.5:

1)

Centrifugal charging pump, a 2480 psid:

l l

2)

Safety Injection pump, a 1445 psid: and i

l 3)

RHR pump. 2 171 psid.

i i

l SEABROOK - UNIT 1 3/4 5-6 Amendment No. 3.

- l i

i EMERGENCY CORE COOLING SYSTEMS i

ECCS SUBSYSTEMS - T GREATER THAN OR E00AL TO 350 F m

l SURVEILLANCE REOUIREMENTS 4.5.2 (Continued) g.

By verifying the correct position of each electrical and/or mechanical position stop for the following~ECCS throttle valves:

1)

Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months following completion of each valve stroking 1

operation or maintenance on the valve when the ECCS subsystems 1

are required to be OPERABLE. and j

2)

At least once per 18 months.

4 Hiah Head SI System Intermediate Head SI System j

Valve Number Valve Number i

SI-V-143 SI-V-80 SI-V-147 SI-V-85 SI-V-151 SI-V-104 SI-V-155 SI-V-109 i

SI-V-117 3

SI-V-121 a

SI-V-125 SI-V-129 i

h.

By performing a flow balance test, during shutdown. following completion of modifications to the ECCS subsystems that alter the subsystem flow characteristics and verifying that:

l i

1)

For centrifugal charging pump lines, with a single pump running:

j a)

The sum of the injection line flow rates. excluding the highest flow rate is greater than or equal to 306 gpm.

[

and i

4 b)

The total pump flow rate is less than or equal to 549 gpm.

l l

j 2)

For Safety Injection pump lines, with a single pump running:

a)

The sum of the injection line flow rates, excluding.the highest flow rate, is greater than or equal to 419 gpm,-

l and b)

The total pump flow rate is less than or equal to 669'gpm.

l 3)

For RHR pump lines. with a single pump running, the sum of the-injection line flow rates is greater than or equal to.4213 gpm.

l i

SEABROOK - UNIT 1 3/4 5-7

' Amendment No. Ei y

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l 3/4.2 POWER DISTRIBUTION LIMITS i

l-BASES I

The specifications of this section provide assurance of fuel integrity i

during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) i events by:

(1) maintaining the minimum DNBR in the core greater than or equal to the design DNBR value during normal operation and in short-term transients.

[

l and (2) limiting the fission gas release, fuel pellet temperature. 'and cladding mechanical properties to within assumed. design criteria.

In addition, limiting i

the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200*F is not exceeded.

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

Fo(Z)

Heat Flux Hot Channel Factor, is defined as the maximum 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:

I FL Nuclear Enthalpy Rise Hot Channel Factor is defined as the ratio of the integral of linear power along the rod with the highest integrated power to the average rod power; and

{

3/4 2.1 AXIAL FLUX DIFFERENCE l

t I

The limits on AXIAL FLUX DIFFERENCE (AFD) specified in the CORE OPERATING LIMITS REPORT (COLR) assure that the design limits on peak local power density and minimum DNBR are not exceeded during normal operation and the consequences of any Non-LOCA event would be within specified acceptance criteria.

i For operation with the Fixed Incore Detectors (FIDS). assurance that the i

F (Z) limit of Specification 3.2.2 is not exceeded during either normal aoperation or in the event of xenon redistribution following )ower changes is L

provided by a separate Fixed Incore Detector Alarm through t1e plant process computer. A FIDS Alarm will be generated when a predetermined number of individual detectors exceed their alarm setpoint. The setpoint for each 1

individual detector is adjusted by the normal 5.21% for system measurement i

uncertainty and 3% for engineering uncertainty. This assures'that the consequences of a LOCA would be within specified acceptance criteria.

Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm. The computer determines the 1-minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for two or more OPERABLE excore channels are outside the limits specified in the COLR.

These alarms are active when power is greater than 50% of RATED THERMAL POWER.

i SEABROOK - UNIT 1 B 3/4 2-1 Amendment No.-9, a

I 4

POWER DISTRIBUTION LIMITS l

BASES I

3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and NUCLEAR ENTHALPY RISE HOT i

CHANNEL FACTOR j

e i

The limits on heat flux hot channel factor and nuclear enthalpy rise hot i

channel factor ensure that:

(1) the design limits on peak local power density and minimum DNBR are not exceeded and (2) in the event of a LOCA, the peak fuel clad temperature will not exceed the 2200 F ECCS acceptance criteria limit.

Each of these is measurable but will normally only be. determined i

periodically as specified in Specifications 4.2.2 and 4.2.3.

This periodic surveillance is sufficient to ensure that the limits are maintained provided:

a. Control rods in a single grou) move together with no individual rod l-insertion differing by more tlan 12 steps, indicated from the group demand position:

b. Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.6:

c. The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 l

are maintained; and

d. The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE.1s maintained within the limits.

1 l

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SEABROOK - UNIT 1 B 3/4 2-2 Amendment No.-

i r

,.e...,.. ~. -

POWER DISTRIBUTION LIMITS-j BASES 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and NUCLEAR ENTHALPY RISE HOT l

CHANNEL FACTOR (Continued)

FL will be maintained within its limits provided Conditions a. through.

d. above are maintained.

l j

The design limit DNBR includes margin to offset any rod bow penalty.

Margin is also maintained between the safety analysis limit DNBR and the design limit DNBR. This margin is available for plant design flexibility.

When an F measurement is taken, an allowance for both measurement error a

J and manufacturing tolerance must be made. An allowance of 5% is appropriate for a full-core map taken with the movable incore-detectors, while 5.21% is appropriate for surveillance results determined with the fixed incore detectors. A 3% allowance is appropriate for manufacturing tolerance.

For operation with the FIDS Alarm OPERABLE. the cycle-dependent normalized axial peaking factor, K(Z), specified in the COLR accounts for axial l

power shape sensitivity in the LOCA analysis. Assurance the F (Z) limit of l

a Specification 3.2.2 is met during 'both normal operation and in the event of xenon redistribution following power changes is provided by.the FIDS Alarm j

through the plant process computer. This assures that the consequences of a l

LOCA would be within specified acceptance criteria.

i j

For operation with the FIDS Alarm inoperable, the cycle-dependent i

normalized axial peaking factor. K(Z), specified in the COLR accounts for

-l possible xenon redistribution following power changes in addition to axial-power shape sensitivity in the LOCA analysis.

This assures that the consequences of a LOCA would be within specified acceptance criteria.

i When RCS FL is measured, no additional allowances are necessary prior to i

comparison with the established limit.

A bounding measurement error of 4.13%

i for FL has been allowed for in determination _of the design DNBR value.

3/4.2.4 OUADRANT POWER TILT RATIO

{

l The purpose of this specification is to detect gross changes in core power distribution between monthly incore detector system surveillances.

During normal operation the OUADRANT POWER TILT RATIO is set equal to zero once 2

acceptability of core peaking factors has been established by review of incore surveillances. The limit of 1.02 is established as an indication that the l

power distribution has changed enough to warrant further investigation.

j i

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SEABROOK - UNIT 1 B 3/4 2-3 Amendment No. 9-14.

i

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POWER DISTRIBUTION LIMITS i

BASES j

i 3/4.2.5 DNB PARAMETERS l

The limits on the DNB-related parameters assure that each of the l

parameters is maintained within the normal steady-state envelope of operation j

assumed in the transient and accident analyses. The limits are consistent with the updated FSAR assumptions and have been analytically demonstrated to assure compliance with acceptance criteria for each analyzed transient. Operating procedures include allowances for measurement and indication uncertainty so that the limits of 594.3 F for T,y and 2185 psig for pressurizer pressure are i

not exceeded.

RCS flow must be greater than or equal to,1) the. Thermal Design Flow l

(TDF) with an allowance for measurement uncertainty and. 2) the minimum measured flow used in place of the TDF in the analysis of DNB related events when the Revised Thermal Design Procedure (RTDP) methodology is utilitized.

The 12-hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.

The periodic surveillance of indicated RCS flow is sufficient to detect only flow degradation which could lead to operation outside the specified limit.

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SEABRDOK - UNIT 1 B 3/4 2-4 Amendment No. 9. 12.

DESIGN FEATURES-DESIGN PRESSURE AND TEMPERATURE 5.2.2 The containment building is designed and shall be maintained for a maximum internal pressure of 52.0 psig and a temperature of 296 F.

j 5.3 REACTOR CORE i

FUEL ASSEMBLIES 5.3.1 The core shall contain 193 fuel assemblies with each fuel assembly i

containing 264 fuel rods clad with a zirconium allny.

Each fuel rod shall have a nominal active fuel length of 144 inches. The initial core loading shall j

have a maximum enrichment of 3.15 weight percent U-235. Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum enrichment of 5.0 weight percent U-235.

CONTROL ROD ASSEMBLIES 5.3.2 The core shall contain 57 full-length control rod assemblies.

The full.

length control rod assemblies shall contain a nominal 142 inches of absorber i

material. The nominal values of absorber material shall be 80% silver,15%

indium, and 5% cadmium. All control rods shall be clad with stainless steel tubing.

i 5.4 REACTOR COOLANT SYSTEM j

DESIGN PRESSURE AND TEMPERATURE 5.4.1 The Reactor Coolant System is designed and shall be maintained:

a. In accordance with the Code requirements specified in.Section 5 2 of the FSAR. with allowance for normal degradation pursuant to the applicable Surveillance Requirements.

j

b. For a pressure of 2485 psig. and i

c. For a temperature of 650 F. except for the pressurizer which is 680 F.

VOLUME 5.4.2 The total water and steam volume of the Reactor Coolant System is 12.265 cubic feet at a nominal T,y of-588.5 F.

5.5 METEOROLOGICAL TOWER LOCATION l

l 5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1.

[

SEABROOK - UNIT 1 5-9 Amendment.No. 6.

y ry y,

yr g.-

y g-c.

7-m.

.- - -.y N

i l

ADMINISTRATIVE CONTROLS 4

i i

ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT l

6.8.1.4 (Continued) i r

to show conformance with 40 CFR Part 190. " Enviro 7ntal Radiation Protection Standards for Nuclear Power Operation." Acceptable methods for calculating the i

dose contribution from liquid and gaseous effluents are given in Regulatory Guide 1.109. Rev. 1. October 1977.

l' The Annual Radioactive Effluent Release Report shall include a list and j

description of unplanned releases from the site to UNRESTRICTED AREAS of radioactive materials in gaseous and liquid effluents made during the reporting

.i period.

l The Annual Radioactive Effluent Release Report shall include any changes made during the reporting period to the PROCESS CONTROL PROGRAM and the ODCM.

pursuant to Specifications 6.12 and 6.13. respectively, as well as any major change to Liquid. Gaseous, or Solid Radwaste Treatment Systems pursuant to Specification 6.14.

It shall also include a listing of new locations for dose l

calculations and/or environmental monitoring identified by the Land use Census pursuant to Specification 3.12.2.

l f

The Annual Radioactive Effluent Release Re) ort shall also include the following: an explanation as to why the inopera)ility of liquid or gaseous

(

effluent monitoring instrumentation was not corrected within the time specified 1

in Specification 3.3.3.9 or 3.3.3.10. respectively: and description of the events leading to liquid holdup tanks or gas storage tanks et eeding the limits of Specification 3.11.1.4 or 3.11.2.6. respectively.

I MONTHLY OPERATING REPORTS 6.8.1.5 Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the U.S. Nuclear Regulatory Commission.

Washington. D.C. 20555. Attn:

Document Control Desk. with a copy to the NRC l

Regional Administrator. no later than the 15th of each month following the.

calendar month covered by the report 1

CORE OPERATING LIMITS REPORT 6.8.1.6.a Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT prior to each reload cycle, or prior to any 2

remaining portion of a reload cycle. for the following:

1.

Cycle dependent Overpower AT and Overtemperature AT trip setpoint parameters and function modifiers for operation with skewed axial power profiles for Table 2.2-1 of Specification 2.2.1.

2.

SHUTDOWN MARGIN limit for MODES 1. 2. 3. and 4 for Specification 3.1.1.1.

3.

SHUTDOWN MARGIN limit for MODE 5 for Specification 3.1.1.2.

4.

Moderator Temperature Coefficient BOL and EOL limits, and 300 ppm surveillance limit for Specification 3.1.13.

SEABROOK - UNIT 1 6-18 Amendment No. 9. 22.

I

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I ADMINISTRATIVE CONTROLS l

6.8.1.6.a. (Continued) 5.

Shutdown Rod Insertion limit for Specification 3.1.3.5.

l 6.

Control Rod Bank Insertion limits for Specification 3.1.3.6.

t 7.

AXIAL FLUX DIFFERENCE limits for Specification 3.2.1.

8.

Heat Flux Hot Channel Factor. F7 and KCZ) for Specification 3.2.2.

l a

9.

Nuclear Enthalpy Rise Hot Channel Factor, and F7 for Specification 3.2.3.

3 l

The CORE OPERATING LIMITS REPORT shall be maintained available in the Control-Room.

i l

l 6.8.1.6.b The analytical methods used to determine the core operating limits 1

shall be those previously reviewed and approved by the NRC in:

l i

1. WCAP-10266-P-A. Rev. 2 with Addenda (Pro]rietary) and WCAP-11524: A l

(Nonproprietary)

"The 1981 Version of t1e Westinghouse ECCS l

Evaluation Model Using the BASH Code". August,1986 l

l Methodology for Specification:

3.2.2 Heat Flux Hot Channel Factor

2. WCAP-10079-P-A (Proprietary) and WCAP-10080-A (Nonpro]rietary).

"NOTRUMP: A Nodal Transient Small Break and General letwork Code".

August. 1985 l

Methodology for Specification:

i 3.2.2 Heat Flux Hot Channel Factor

3. YAEC-1363-A. "CASMO-3G Validation.'" April 1988.

YAEC-1659-A " SIMULATE-3 Validation and Verification." September 1988.

Methodology for Specifications:

3.1.1.1 SHUTDOWN MARGIN for MODES 1. 2. 3. and 4 3.1.1.2 -

SHUTDOWN MARGIN-for MODE 5 3.1.1.3 Moderator Temperature Coefficient j

3.1.3.5 -

Shutdown Rod Insertion Limit S

3.1.3.6 -

Control Rod. Insertion Limits -

I i

3.2.1 AXIAL FLUX DIFFERENCE 3.2.2 Heat Flux Hot Channel Factor 3.2.3 Nuclear Enthalpy Rise Hot Channel Factor

.I l

4. Seabrook Station Updated Final Safety Analysis Re) ort. Section~ 15.4.6.

" Chemical and Volume Control System Malfunction Tlat Results in a Decrease in the Boron Concentration in the Reactor Coolant System".

Methodology for Specifications:

3.1.1.1 SHUTDOWN MARGIN for MODES 1. 2. 3 and 4 3.1.1.2 -

SHUTDOWN MARGIN for MODE 5 SEABROOK - UNIT 1 6-18A Amendment No. 9.

ADMINISTRATIVE CONTROLS f

l 6.8.1.6.b. (Continued)

5. YAEC-1241. " Thermal-Hydraulic Analysis of PWR Fuel Elements Using the CHIC-KIN Code". R. E. Helfrich. March 1981 5

Methodology for Specification:

3.2.1 AXIAL FLUX DIFFERENCE i

3.2.2 Heat Flux Hot Channel Factor 3.2.3 Nuclear Enthalpy Rise Hot Channel Factor j

6. YAEC-1849P, " Thermal-Hydraulic Analysis Methodology Using VIPRE-01 For PWR Applications. " October 1992 Methodology for Specification:

r 2.2.1 Limiting Safety System Settings 3.2.1 AXIAL FLUX DIFFERENCE 3.2.2 Heat Flux Hot Channel Factor 3.2.3 Nuclear Enthalpy Rise. Hot Channel Factor

/. YAEC-1854P. " Core Thermal Limit Protection Function Setpoint i

Methodology For Seabrook Station. " October 1992 i

i Methodology for Specification

~

2.2.1 Limiting Safety System Settings 3.1.3.5 -

Shutdown Rod Insertion Limit 3.1.3.6 -

Control Rod Insertion Limits 3.2.1 AXIAL FLUX DIFFERENCE 3.2.2 Heat Flux Hot Channel Factor 3.2.3 Nuclear Enthalpy Rise Hot Channel Factor 8

YAEC-1856P. " System Transient Analysis Methodology Using RETRAN for i

PWR Applications." December 1992 Methodology for Specification:

2.2.1 Limiting Safety System' Settings 3.1.1.3 -

Moderator Temperature Coefficient 3.1.3.5 -

Shutdown Rod Insertion Limit 3.1.3.6 -

Control Rod Insertion Limits 3.2.1 AXIAL FLUX DIFFERENCE 3.2.2 Heat Flux Hot Channel Factor 3.2.3 Nuclear Enthalpy Rise Hot Channel Factor-9.

YAEC-1752. " STAR Methodology A) plication for PWRs. Control Rod l

Ejection. Main Steam Line Breac," October 1990 l

Methodology for Specification.

3.1.1.3 -

Moderator Temperature Coefficient 3.1.3.5 -

Shutdown Rod Insertion Limit 3.1.3.6 -

Control Rod Insertion Limits 3.2.1 AXIAL FLUX DIFFERENCE 3.2.2 Heat Flux Hot Channel Factor 3.2.3 Nuclear Enthalpy Rise Hot Channel Factor SEABROOK - UNIT 1 6-188

. Amendment-No.- 9.

u - -a

...m.,

I r

l ADMINISTRATIVE CONTROLS 6.8.1.6.b. (Continued) 10.

YAEC-1855P, "Seabrook Station Unit 1 Fixed Incore Detector System Analysis." October 1992 j

Methodology for Specification:

3.2.1 AXIAL FLUX DIFFERENCE 3.2.2 Heat Flux Hot Channel Factor i

3.2.3 Nuclear Enthalpy Rise Hot Channel Factor 11.

YAEC-1624P. " Maine Yankee RPS Setpoint Methodology Using Statistical Combination of Uncertainties - Volume 1 - Prevention of Fuel r

Centerline Melt." March 1988 Methodology for Specification.

3.2.1 AXIAL FLUX DIFFERENCE i

3.2.2 Heat Flux Hot Channel Factor 3.2.3 Nuclear Enthalpy Rise Hot Channel Factor l

i 6.8.1.6.c.

The core operating limits shall be determined so that all l

a)plicable limits (e.g., fuel thermal-mechanical limits, core'.

i tiermal-hydraulic limits. ECCS limits, nuclear limits such as SHUTDOWN MARGIN.

~

and transient and accident analysis limits) of the safety analysis are met.

The CORE OPERATING LIMITS REPORT for each reload cycle, including any mid-cycle-revisions or supplements thereto, shall be provided upon issuance, to the NRC j

l Document Control Desk with copies to the Regional Administrator and the L

Resident Inspector.

s i

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SEABROOK - UNIT 1 6-18C Amendment No. 9.

IV.

Safety Evaluation of Liernse Amendment Request 93-09 Proposed Chances The purpose of License Amendment Request 93-18 is to propose changes to the Seabrook Station Technical Specifications to allow North Atlantic to operate the core within an axial flux difference (AFD) band which is expanded from the current band in the Technical Specifications. The proposed changes will also allow improved fuel cycle management through the implementation of care design enhancements. The proposed expanded AFD band and core design enhancements include modification to several safety analysis input parameters and assumptions. These are:

Incorporation of the Westinghouse WRB-1 Depanure from Nucleate Boiling (DNB)

Correlation and Revised Thermal Design Procedure (RTDP)

Increase in the core power distribution peaking factors Allowance for positive moderator temperature coefficient Allowance for thimble plug deletion Allowance for increase in steam generator tube plugging

+

Flexibility to implement certain new fuel design features in the future. e.g., low pressure drop Zircalloy grids and Zirlo cladding Modification of analysis assumptions related to certain surveillance parameters, e.g., low pressurizer pressure safety injection actuation setpoint and time delay Expansion of the AFD Limiting Condition for Operation (LCO) band The analytical methods used were dec sed for Nonh Atlantic be Yankee Atomic Electric Company (Y AEC) and are consistent with hdC approved methods applied by YAEC at the Maine Yankee and Yankee Rowe plants. Minor changes in the approved methods currently used were j

necessary to accommodate specific design features at Seabrook Station. The specific design i

features at Seabrook Station for which the methods were tailored are automatic rod motion and j

the form of the thermal protection functions. The specific methodologies used at Seabrook Station were previously submitted for NRC review in February 1993 in preparation for the submittal of this License Amendment Request [ Reference (b)].

The methodology used includes consideration for the use of the Fixed Incore Detectors (FIDS) to continuously monitor compliance of the core power distribution with the fechnical Specification limits imposed by the Loss of Coolant Accident (LOCA) analysis assumptions.

Compliance with the LOCA limits is currently assured by operation within an Axial Flux Difference (AFD) band known as Constant Axial Offset Control (CAOC). In the CAOC monitoring methodology, the excore detector axial flux difference indications are used to infer worst case core power distributions. Core operation is limited to an AFD envelope based on worst case power distributions pre-determined by conservative off-line physics analysis.

Use of the FIDS eliminates much of the uncertainty in quantifying available margin to the LOCA limits associated with off-line inference of worst case power distributions from the excore signals.

This allows the AFD Limiting Condition for Operation (LCO) to be expanded beyond that specified by CAOC philosophy. The AFD LCO band becomes defined by power distribution 8

l restraints required to ensure adequate initial margin to fuel thermal design limits on departure from nucleate boiling (DNB) and fuel centerline meh for anticipated transients.

For operation with the FIDS Alarm inoperable, the cycle dependent normalized axial peaking factor, K(Z), specified in the COLR, accounts for possible xenon redistribution following power changes in addition to axial power shape sensitivity in the LOCA analysis. This ensures that the consequences of a LOCA would be within the specified acceptance criteria.

A safety analysis in support of operation with an expanded axial flux difference Limiting Condition for Operation (LCO) band and enhanced core and system design features is provided in YAEC-1871, " Safety Analysis in Support Wide Band Operation and Core Design Enhancements for Seabrook Station". (Enclosure 1) l Justification for the expanded axial slux difference LCO band and the enhanced design core parameters is provided through a complete re-analysis of the Seabrook Station Updated Final Safety Analysis Report (UFSAR) Chapter 15 Accidents and Transients. The large and small break l

LOCA analysis was performed for North Atlantic by Westinghouse Electric Corporation j

(Enclosure 2). The remaining Accidents and Tunsients of the UFSAR Chapter 15 were evaluated l

for North Atlantic by YAEC.

l The results of the safety analysis demonstrate that Seabrook Station can be safely operated within i

the Technical Specifications including the proposed changes. Therefore, since the proposed changes will continue to operate the plant within the bounds of the Technical Specifications there i

is no increase in the safety consequences associated with the requested amendment.

l I

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9

V.

Determination of Sinnificant llazards for Iicense Amendment Request 93-09 Proposed Chances 1.

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

The changes proposed in License Amendment Request (LAR) 93-18 involve modifications to assumptions in the safety analysis and do not change the manner that systems or components operate from that previously evaluated. LAR 93-18 proposes changes to the manner in which power distributions are controlled, but the basic criteria used to define the allowable operating envelope of power distributions are not changed. Therefore, the probability of an event previously evaluated has not increased by the proposed changes.

Each accident and transient identified in the Seabrook Station Updated Safety Analysis Report (UFSAR) has been evaluated for the proposed changes. The results of this evaluation have been documented in YAEC-1871, " Safety Analysis In Support of Wide-Band Operation and Core Design Enhancements for Seabrook Station" and in "Seabrook Station Fuel Upgrade Program LOCA Safety Analysis Report" The conclusion from these evaluations is that all acceptance criteria have been met including radiological consequences. Since the plant response to an accident will not change there is no change in the potential for an increase in the release of radiation to the public from the proposed changes. Therefore, it follows that the consequences of a previously evaluated accident, as measured in terms of dose, will not significantly increase due to the proposed changes.

(2)

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

As described above, the changes proposed by this License Amendment Request involve changes to assumptions in the safety analysis and do not change the manner in which systems or components operate from that previously evaluated. In License Amendment Request 92-14, "Incore Detector System" [ Reference (c)], North Atlantic proposed using the Fixed Incore Detectors (FIDS) as an independent system to perform Technical Specification Surveillances.

The use of the FIDS eliminates much of the uncertainty in quantifying avr.ilable margin to the L-OCA limits associated with off-line inference of worst case power distributions from the excore signals. This allows the Axial Flux DifTerence (AFD) Limiting Condition for Operation (LCO) to be expanded beyond that specified by the Constant Axial Offset Control (CAOC) philosophy.

The AFD LCO band becomes defined by power distribution restraints required to ensure adequate initial margin to fuel thermal design limits on departure from nucleate boiling (DNB) and fuel centerline melt for anticipated transients. Use of the FIDS alarm through the plant process computer provides assurance that the F (Z) limit of Technical Specification 3.2.2, IIeat Flux Ilot n

Channel Factor will not be exceeded during both normal operation and in the event of a xenon redistribution following power changes. This ensures that the consequences of a LOCA would be within the specified acceptance criteria.

For operation with the FIDS alarm inoperable, the cycle dependent normalized axial peaking factor, K(Z), specified in the COLR, accounts for possible xenon re istribution following power d

changes in addition to axial power shape sensitivity in the LOCA analysis. This assures that the consequences of a LOCA would be within the specified acceptance criteria.

10

~

The changes either involve modifications to input assumptions in the safety analysis or use of the FIDS to provide continuous detailed information on core power distributions. There are no modifications to the manner in which plant equipment operates associated with the proposed changes and no new failure mechanisms are introduced. Therefore, the proposed changes do not create the possibility of a new or different accident from any accident previously evaluated.

(3)

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

l The transient and accident analysis presented in the Seabrook Station UFSAR defines the margin of safety in the acceptance criteria specified for each event. Each accident and transient identified in the UFSAR has been evaluated for the changes in the following parameters:

Incorporation of the Westinghouse WRB-1 Departure from Nucleate Boiling (DNB)

Correlation and Revised Thermal Design Procedure (RTDP)

Increase in the core power distribution peaking factors Allowance for positive moderator temperature coefficient Allowance for thimble plug deletion

+

Allowance for increase in steam generator tube plugging Flexibility to implement certain new fuel design features in the future. e.g., low pressure drop Zircalloy grids and Zirlo cladding Modification of analysis assumptions related to cenain surveillance parameters, e.g., low pressurizer pressure safety injection actuation setpoint and time delav Expansion of the AFD LCO band i

+

A safety analysis in suppon of operation with an expanded axial flux difference Limiting Condition for Operation (LCO) band and enhanced core and system design features is provided in YAEC-1871, " Safety Analysis in Suppon Wide Band Operation and Core Design Enhancements for Seabrook Station". (Enclosure 1)

Justification for the expanded axial flux difference LCO band and the enhanced design core parameters is provided through a complete re-analysis of the Seabrook Station Updated Final Safety Analysis Report (UFSAR) Chapter 15 Accidents and Transients. The large and small break LOCA analysis was performed for North Atlantic by Westinghouse Electric Corporation.

(Enclosure 2). The remaining Accidents and Transients of the UFSAR Chapter 15 were evaluated i

for Nonh Atlantic by YAEC.

The DNB evaluations have applied the Westinghouse WRB-1 DNB correlation and the Revised Thermal Design Procedure methodology. This methodology provides a new acceptance limit which is closer to DNB than the previously utilized the Westinghouse W-3 DNB correlation. The new acceptance limit is a result of a more accurate determination of predicted DNB. This affects the margin of safety as defined in the Bases of the Technical Specifications.110 wever, the DNB design basis remains the same, specifically, that there must be at least a 95% probability at a 95%

confidence level that the limiting power fuel rod will not experience DNB during normal 11

._=

I

)

s r

operation and anticipated operational occurrences. The Westinghouse WRB-1 DNB correlation and RTDP have been approved by the NRC for use on Westinghouse fuel of the design used at Seabrook Station. Furthennore, the revised parameters proposed in LAR 93-18 are consistent with current industry practice.

The results of the enclosed safety analysis demonstrate that Seabrook Station can be safely operated within the Technical Specifications including the proposed changes. Therefore, the proposed changes do not result in a significant reduction in the margin of safety.

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VI.

Proposed Schedule for License Amendment Issuance and Effectiveness North Atlantic requests NRC review of License Amendment Request 93-18 and issuance of a license amendment having immediate effectiveness by July 15.1994.

a b

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VII.

Envirtmmental Impact Assessment North Atlantic has reviewed the proposed license amendment against the criteria of 10CFR51.22 for environmental considerations. The proposed changes do not involve a significant hazrds consideration, nor increase the types r.nd amounts of effluents that may be released ofTsite, r.or significantly increase individual or cumulative occupational radiation exposures. Based on the foregoing, North Atlantic concludes that the preposed change meets the criteria delineated in 10CFR51.22(c)(9) for a categorical exclusion from the requirements for an Environmental Impact Statement.

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VIII. Supportine Information YAEC-1871 " Safety Analysis in Support of Wide-Band Operation and Core Design Enhancements for Seabrook Station" Seabrook Station Fuel Upgrade Program LOCA Safety Analysis Report Proposed Core Operating Limits Report 10CFR50.46 LOCA Model Assessments on the PCT Margin Utilization c

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3.2. ACTION

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