Text

Amend 12 to License NPF-86,providing Replacement of RTD Bypass Sys for Measurement of Primary Loop Temp w/fast-response thermowell-mounted RTDs
	ML20099G298
Person / Time
Site:	Seabrook
Issue date:	08/10/1992
From:	; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20099G299	List: ML20099G298;
References
	NUDOCS 9208140194
	Download: ML20099G298 (19)
	v • d • e

'

$~g,, n g%

.:' %' D

'n UNITED STATES E:

E NUCLEAD REGULATORY COMMISSION

.f WASHINGTON. D.C 20%S e,

f.

....+

30RTH ATLANTIC ENERGY SERVICE CORPORATION. ET AL.*

DOCKET NO. 50-443 MABROOK STATIIL_ UNIT NO.1 M[2AENT TO FACILITY OPERATlhG LICENSE Amendment No.12 License No. NPF-86 1.

The Nuclear Regulatory Commission (the Commission or the NRC) has found that:

A.

The application for amendment filed by the North Atlantic Ei.ergy Service Corporation (NAESCO) f the licensee), acting for itself and as agent anj representative of the 11 other utilities listed below and hereaf',er referred to as licensees, dated March 20, 1992, as supphmented on June 19, 1992, complies with the standatds and requirements of the Atomic Energy Act of 1954, as amended (the Act),

and the Commission's rules and regulations set forth in 10 CFR Chapter I;

B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is rnasonable assurance- (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's reg'dation and all applicable requirements have been satisfied.

North Atlantic Energy Service Corporation is authorized to act as agent for the North Atlantic Energy Corporation, the Canal Electric Company, The Connecticut Light and Power Company, EUA Power Corporation, Hudson Light &

Power Department, Massachusetts Municipal Wholesale Electric Company, Montaup Electric Company, New England Power Company, New Hampshire Ehetric Cooperative, Inc., Taunton Municipal Light Plant, The United Illuminating Company, and Vermont Electric Generation and Transmission Cooperative, Inc.,

and has exclusive responsibility and control over the physical construction, operation and maintenance of the facility.

9208140144 920810 DR ADCCK 0500 3

l 2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of facility Operating License No. NPF-86 is hereby amended to read as follows:

(2) Technical Specificttions The Technical Specifications contained in Appendix A, as revised through Amendment No.12, and the Environmental Protection Plan contained in Appendix 8 are incorporated into Facility License No.

NPF-86.

NAESCO shall operate the facility in accordance with the Technical Specification > and the Environment,- Protection Plan.

3.

This license amendment is effectne as of its date of issuance.

The upgrades and enhancements associated with replacement of the RTD Bypass System in this amendment will be implemented prior to entry into Mode 3 during restart from the second eefueling outage.

The resistance tiermal detector cross-calibration and rasponse time tests, and a reactor coolant system leak test can be completed following entry into Mode 3.

Additi<n-ally, a flow calorimetric measurement will be performed upon chieving stable full power,eration. All other testing required to demonstrate proper operation of modified compan!nts will be completed prior to entry into Mode 3.

FOR THE NUCLEAR REGULATORY COMMISSION N

Victor Herses, Acting Director Project Directorate 1-3 Division of Reactor Proje-'s - 1/11 Office of Nuclear Reactor. gulation

Attachment:

Changes to the Technical Specifications Date of issuance:

August 10, 1992 1,

o

. - --.-.~. -.

ATTACHMENT TO LICENSE AMENDMENT NO. 12 FACIllTY OPERATING LICENSE NO. NPF-86 DOCKET NO.59-14J l

Replace the following pages of the Appendix A Technical Specifications with the attached pages.

The revised pages are identified by Amendment number and contair vertical lin1s indicating the area of change.

Overlap pages have been provided.

Remove Insert 2-4 2-4 2-5 2-5 2-7 2-7 2-8 2-8 2-10 2-10 3/4 2-10 3/4 2-10 3/4 3-9 3/4 3-9 3/4 3-13 3/4 3-13 8 3/4 2-4 B 3/4 2-4 n

- +

,-,--,e

-y-a-.-

-eo w

.--m.

e+-

,,mw---vr--,-,,--.,-,--ww--

, -,,.,r-.-.,-,--m.-.m---+,,

atw cww.rwy---,-,m y, sw + -- ww.e -e-e y v-w e e--

em.---w,---r-,rr-w-r-v

~.

i e'

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The Reactor Trip System Instrumentation and Interlock Setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2-1.

APPLICABILITY:

As shown for each channel in Table 3.3-1.

ACTION:

With a Reacter Trip System Instrumentation or Interlock Setpoint a.

less conservative th6n the value shown in the Trip Setpoint column but.more conservative than the value shown in the Allowable Value column of Table 2.2-1, adj-:t the Setpoint consistent with the Trip Setpoint value.

b.

With the Reactor Trip System Instrumentation or Interlock Setpoint less conservative than the value shown in the Allowable Value column of Table 2.2-1, either:

1.

Adjust the Setpoint consistent with the Trip Setpoint value of Table 2.2-1 and determine within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that Equation 2.2-1 was satisfied for the affected channel, or 2.

Declare the channel inoperable and apply the appilcable ACTION statement. requirement of Specification 3.3.1-until the channel is restored to OPERABLE status with its Setpoint adjusted consistent with the Trip Setpoint value.

Equation 2.2-1 Z + R + 5 < TA Where:

Z = The value from Column Z of Table 2.2-1 for the affected

channel, R = The "as measured" value (in percent span) of rack error for the affected channel, S = Either the "as measured" value (in percent span) of the sensor error, or the value from Column 5 (Sensor Error) of Table 2.2-1 for the affected channel, and TA = The value from Column TA (Total Allowance) of Table 2.2'1 f

for the affected channel.

SEABROOK - UNIT 1 2-3

TABLE 2.2-1 M

8; REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS SENSOR x

TOTAL ERROR FUNCTIONAL UNIT ALLOWANCE (TA)

Z (5)

TRIP SETPOINT ALLOWABLE VALUE

'1.

Manual Reactor Trip N.A.

N.A.

~

2.

Power Range, Neutron Flux a.

High Setpoint 7.5 4.56 0

<109% of RTP*

$111.1% of RTP*

b.

Low Setpoint 8.3 4.56 0

$25% of RTP*

$27.1% of RTP*

3.

Power Ranoe, Neutron Flux,

1. 6 0.5 0

<5% of RTP* with

<6.3% of RTP* with High Posi;ive Rate i time constant i *.ime constant 12 seconds 12 seconds 4.

Power Range, Neutron Flux, 1.6 0.5 0

<5% of RTP' with

<6.3% of RTP* with y

High Negative Rate i time constant i time constant 22 seconds 32 seconds

=

5.

Intermediate Range, 17.0 8.41 0

$25% of RTP*

$31.1% of RTP*

Neutron Flux 5

6.

Source Range, Neutron Flux 17.0 10.01 0

$10 cps

$1.6 x 105 cps 7.

Overtemperature AT 6.5 3.5 1.7**

See Note 1 See Note 2 i

+0.5**

I if 8.

Overpower AT 4.9 2.2

1. 7 See Note 3 See Note 4

]

g 9.

Pressurizer Pressure - Low 3.12 0.86 0.99 11945 psig 11,931 psig g 10.

Pressurizer Pressure - High 3.12 1.00 0.99

$2385 psig

<2,398 psis

RTP = RATED THERMAL POWER

- The sensor errar for T,yg is 1.7 and the sensor error for Pressurizer Pressure is 0.5.

"As measured" sensor errors my be used in lieu of either or both of these values, which then must be sumed to deter-mine the overtemperature AT total channel value for S.

O

TABLE 2.2-1 (Continued) v,

!5 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 8

SE SENSOR TOTAL ERROR' FUNCTIONAL UNIT ALLOWANCE (TA)

Z (S)

TRIP SETPOINT ALLOWABLE VALUE c:5

-4 11.

Pressurizcr Water Level - High 8.0 4.20 0.84

$92% of instrument

$93.75% of instrument span span 12.

Reactor Coolant Flow - Low 2.5 1.9 0.6

>90% of loop

>89.3% of loop design flow'-

design flow *

13. Steam Generator Water 14.9 12.53 0.55

>14.0% of narrow

>12.6% of narrow Level Low - Low range instrument range instrument span span 14.

I'ndervoltage - Reactor 15.0 1.39 0

>10,200 volts

>9,822 mits in"lant Pumps 15.

Underfrequency - Reactor 2.9 0

0

>55.5 Hz

>55.3 Hz 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.

y Closure

->l% open

>1% open m

17.

Safety Injection Input N.A.

N.A.

j from ESF r

5 M

Loop design flow = 95,700 gpm

TABLE 2.2-1 (Cont.inued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS E8 SENSOR TOTAL ERROR FUNCTIONAL UNIT Alt 0WANCE (TA)

Z (S)

TRIP SETPOINT All0WABLE VALUE 5

18.

Reactor Trip System Interlocks a.

Intermediate Range N.A.

N.A.

>l x 10 8"

amp ef x 10 28 am;>

Neutron Flux, P-6 b.

Low Power Reactor Trips Block, P-7

1) P-10 input N.A.

N.A.

$10% of R1P*

$12.1% of RTP*

2) P-13 input N.A.

N.A.

<10% PTP* Turbine

<12.3% RTP* Turbine Impul.e Pressure impulse Pressure m

Equivalent Equivalent c.

Pcwer Range Neutron N.A.

N.A.

$50% of RTP*

$52.1% of RTP*

' Flux, P-8 d.

Power Range Neutron N.A.

N.A.

$20% of RTP*

$22.1% of RTP*

Flux, P-9 e.

Power Range Neutron N.A.

N.A.

N.A

>10% of RTP*

>7.9% of RTP*

Flux, P-10 f.

Turbine Impulse Chamber N.A.

N.A.

$10% RTP* Turbine

$17.3% RTP* Turbine Pressure, P-13 Impulse Pressure Impulse Pressure Equivalent Equivalent 15.

Reactor Trip Breakers N.A.

N.A.

N.A N.A.

N.A.

20.

Automatic Trip and Interlock N.A.

N.A.

Logic

RTP = RATED THERMAL POWER s

^

TABLE 2.2-1 (Continued) wg TABLE NOTATIONS

%8 NOTE 1:

OVERTEMPERATURE AT ji

[T

- T 1 + K (P - P') - f (aI)}

]

aT 7

gy T,3) $ AT, Mi -K 3

i 2,1 3

(1 tsS) 5

-4 Measured AT by RTD Instrumentation; l

Where:

Al

=

f d5 I

lead-lag compensator on measured AT;

=

Time constants utilized in lead-lag compensator for AT, 1: >8s,

=

Is. I2 r2 1 3 S; yf 3

Lag compensator an measured AT;

=

au r3 Time constants utilized in the lag canpensater for AT, r3 = 0 s;

=

O Indicated AT at RATED THERMAL POWER; AT,

=

1.0995; K

=

i l

0.0112/ F; K

=

2 1+T5 The function generated by the lead-lag compensator for T l

4

=

avg I*Ib5 dynamic compensation; g

<aR 14, r3 Time constants utilized in the lead-lag compensator for T8*9, 1

_ 33 s,

=

4 M

rs < 4 s; 5

Average temperature, *F; T

=

2 1

Lag compensator on measured Tavg;

=

1 + 1s5

= Tire constant utilized in the measured T,yg lag compensator, is = 0 s; rs t

R TABLE 2.2-1 (continued) gl; TABLE NOTATIONS E

R NOTE 1:

(Continued) e T'

588.5"F (Nominal T,yg at RATED THERMAL POWER);

K-

=

3 0.000519/psig; w

P

=

Pressurizer pressure, psig; P'

2235 psig (Nominal RCS operating pressure);

=

l S

=

Laplace transform operator, s 1;.

I I

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

power-range neutron ion chambers; with gains to be selected based on measured instrument response during plant stari.up tests so that:

ty (1). For qt - gb between - 35% and + 8%, f (AI) = 0, where qt ""d 'b are percent RATED THERMAL 2

POWER in the top and bottom halves of the core respectively, and qt*Ab is total TH N POWER in p=rcent of RATED THERMAL POWER; (2) For each percent that the magnitude of q

~9 exceeds - 35%, the aT Trip Setpoint shall t

b be automatically reduced by 1.09% of its value at RATED IHEPMAL POWER; and

[

(3) For each percent that the magnitude of q ~9b exceeds + 8%, the AT Trip Setpoint shall t

h be automatically reduced by 1.00% of its value at RATED THERMAL POWER.

Eo NOTE 2:

The channel's maximum Trip Setpoint shall not exceed its c m uted Trip Setpoint by more than 2.5%

g of AT span.

D

~.?

TABLE 2.2-1 (Continued)

{

TABLE NOTATIONS (Continued) 8 O

7

. NOTE 3:,

OVERPOWER AT g

37 (1 + tsS)

(1)

(T25)

(1)

(I)

~

q (1 + T 5) (1 + T 5).< ATo {K 2

3 (1 + 1 5) (1 + TsS) T - K [T (1 + tsS) - I"] - f (aI)}

-K 4

5 2

7 As defined in Note 1, Where:

AT

=

f As defined in Note 1,

=

13, T2 As defined in Note 1,

=

i yf 3

As defined in Note 1,

=

7 T3 As defined in Note 1,

=

e 4T, As defined in Note 1

=

1.09, K

=

4 0.02/*F for increasing average temperature and 0 fo: decreasing average Ks

=

temperature, y[7 The function generated by the rate-lag compensator for T,yg dynamic

=

3 compensation, Time constants utilized in the rate-lag compensator for Tavg,17 3 10 s,

=

-17 As defined in Note 1,

=

1 + tsS is As defir.ed in Note 1,

=

t

=

TABLE'2.2-l'(Continued) vs Oh TABLE NOTATIONS (Continued)

U$9 NOTE 3:

(Continued)

C RE' Kc 0.001386/*F; for T > T" and Xs = 0 for T 5. T",

=

w 6'

T As defined in Note 1,

=

T"

=

Indicated T,yg at RATED THERMAL POWER (Calibration temperature for AT instrumentation, i L88.5*F),

As defined in Note 1, and S

=

f (AI) 0 for all AI.

=

2

?

NOTE 4:

The channel's maximum Trip Setpoint shall not exceed its computed 1 rip Setpoint by more than 2.0% of AT span.

}

E 8

an O.

1

L-)

POWER DISTRIBUTION LIMITS 3/4.2.4 QUADRANT POWER T,ILT RATIO LIMITINC CONDITION FOR OPERATION 3.2.4 The QUADRANT POWER TIL1 RATIO shall not exceed 1.02.

APPLICABILITY:

MODE 1 above 50% of RATED THERMAL POWEP*.

ACTION-With the QUADRANT POWER TILT RATIO determined to txceed 1.02:

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

POWER for each 1% of indictted 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 .

b.

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

F evaluation) and F are within their limits by performing Surveil-xy g

lance Requirements 4.2.2.2 and 4.2.3,2.

THERMAL '0WER and setpoint red"0tions shall then be in accordance with the ACTION statements of Specift: ' tons 3.2.2 and 3.2.3.

SURVEILLANCE REQUIREMENTS 4.2.4.1 The QUADRANT POWER TILT RATIO shall be determined to be within the limit above 50% of RATED THERMAL POWER by:

a.

Calculating the ratio at least once per 7 days when the alarm is OPERABLE, and 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.

4.2 4.2 The QUADRANT 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 movable incore detectors to conficm indicated QUADRANT 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 thimble locations or b.

Using the movable incore detection system to monitor the QUADRANT POWER TILT RATIO subject to the requirements of Specification 3.3.3.2.

[

See Special Test Exceptions Specification 3.10.2.

L SEABROOK - UNIT 1 3/4 2-3

)

POWER '11STRIBUTION LIMITS 3/4.2.5 DNB PARlFETERS-LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB-related parameters shall be maintained within the the follow

limits

Reactor Coolant System T,yg, < 594.3*F a.

b.

Pressurizer Pressure: > 2205 psig*

c.

Reactor Coolant System flow, > 392,000 gpm**

APPLICABILITY:

MODE 1.

ACTION:

With any of the abuve 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 o less than 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE 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 fuel loading.

The provisions of Specification 4.0.4 are not applicable for entry into MODE 1.

l

Limit not. applicable durin@ either a-THERHAL 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.

- Includes a 2.4% flow measurement uncertainty.

SEABROOK - UNIT 1-3/4 2-10 Amendment No. 12

1 M

TABLE 4.3-1 h

REACTOR TRIP SYSTEM INSTRUMf'!TATION SURVEILLANCE REQUIREMENTS

-8 TRIP ANALOG ACTUATING MODES FOR E

CHANNEL DEVICE WHICH Z

CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCI FUNC1IONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRE ^..

g 1.

Manual Reactor Trip N.A.

N.A.

R(13)

N.A.

1, 2, 3 *,

2.

Power Range, Neutron Flux a.

High Setpoint S

D(2,4),

Q(16)

N.A.

1, 2 M(3, 4),

(

Q(4, 6),

R(4,5) i b.

Low Setpoint 5

R(4)

S/U(1) ii. A.

N.A.

1***, 2 3.

Power Rangi Neutron Flux, N.A.

R(4)

Q(15)

N.A.

1, 2 1

High Positive Rate m

b 4.

Power Range, Neutron Flux, N.A.

R(4)

Q(16)

N.A.

1, 2 High Negatite Rate 5.

Intermediate Range, S

R(4, 5)

S/U(1)

N.A.

1***, 2 Neutron Flux 6.

Source Range, Neutron Flux 5

R(4, 5)

S/U(1),Q(9,16)

N.A.

2**,3,4,S 4

7.

Overtemperature AT S

R Q(16)

N.A.

1, 2 l

2 8.

Overpowcr AT S

R Q(16)

N.A.

1, 2 9.

Pressurizer Pressure--Low S

R Q(16,17)

N. A.

N.A.

1 g 10.

Pressurizer Pressure--High S

R Q(16,17)

N.A.

1, 2 M 11.

Pressurizer Water Level--High 5 R

Q(16)

N.A.

1 12.

Reactor Coolant Flow--Low 5

R Q(16)

N.A.

I 1

)

i

t TABLE 4.3-1 (Continued) h REACTOR TRIP SYSTEM INSTRUMENTATION S~JRVEILLANCE REQUIREMENTS E

8 TRIP

^

ANALOG ACTUATING MODES FOR CHANNEL DEVILE WHICH t

E CHANNEL.

CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE Z FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED 1

~

i 13.

Steam Generator Water Level-- S R

Q(16, 17)

N.A.

H.a.

i, 2 i

Low-Low 14.

Undervoltage - Reactor Coolant N.A.

R N.A.

Q(16)

N.A.

1 Pumps i

15.

Underfrequency - Reactor N.A.

R N.A.

Q(16)

N.A.

1 Coolant Pumps R

16.

Turbine Trip a.

Low Fluid Oil Pressure N.A.

R N. A.

S/U(1, 10)

N.A.

1 y.

j E;

b.

Turbine Stop Valve N.A.

R N.A.

S/U(1, 10)

N A.

1 Closure i

17.

Safety Injection Input from H.A.

N.A.

R N.A.

1, 2 ESF 18.

Reactor Trip System Interlocks t

a.

Intermediate Range j

Neutron Flux, P-6 N.A.

R(4)

R N.A.

N.A.

2**

b.

Low Power Reactor f

I Trips Block, P-7 N.A.

R(4)

R N. A.

N.A.

1 c.

Power Range Neutron Flux, P-8 N.A.

R(4)

R N.A.

N.A.

I d.

Power Range Neutron Flux, P-9 N.A.

R(4)

R N.A.

M.A.

I 1

l i

l s

TABLE 4.3-1(Continued}

TABLENOTATIONS(Continued}

(12) Number not used.

(13) The TRID ACTUATING DEVICF OPERATIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip circuits for the Manual Reactor Trip Function.

The test shall also verify the OPERABILITY of the Bypass Breaker trip circuit (s).

(14) local manual shunt trip prior to placing breaker in service.

(15) Automatic undervoltage trip.

(16) Each channel shall be tested at least every 92 days on a STAGGERED TEST BASIS.

(17) These channels also provide inputs to ESFAS.

Co. 1 MODES and surveillance frequencies of Specificatio;ry with the applicable

4. 3. 2.1 for any por-tion of the channel required to be OPERABLE by Specification 3.3.2.

l SEABROOK - UNIT 1 3/4 3-13 Amendment No. 12

INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETO FEATURES ACTUATION SYSTEM INSTRUMENTATION i

LIMITmG CONDITION FOR OPERATION 3.3.2 The Engineered Safety Features, Actuation System (ESFAS) instrumentation i

channels and interlocks shown in fable 3.3-3 shall be OPERABLE with their Trip 5etpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4.

APPLICABILITY:

As shown in Table 3.3-3.

ACTION:

With an ESFAS Instrument tion or Interlock Trip Setpoint trip less a.

conservative than the value shown in the Trip Setpoint column but more conservative than the value shown in the Allowable Value column of Table 3.3-4, adjust the Setpoint consistent with the Trip Setpoint value.

b.

With an ESFAS Instrumentation or Interlock Trip Setpoint less conserva-tive than the value shown in the Allowable Value column of Table

3. 3-4, either:

1.

Adjust the Setpoint consistent with the Trip Setpoint value of Table 3.3-4, and determine within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that Equation 2.2-1 was satisfied for the affected channel, or 2.

Declare the channel inoperable and apply the applicable AC1 ION statement requirements of Table 3.3-3 until the channel is restored to OPERABLE status with its Setpoint adjusted consistent with the_ Trip Setpoint value.

Equation 2.2-1 Z + R + 5 < TA Where:

Z-= The value from Coltmn Z of Table 3,3-4 for the affected chanr,el, R = The "as measured" value (in percent span) of rack error for the af fected chanr:el, S = Either the "as measured" value (in percent span) of the sensor error, or the value from Column S (Sensor Error) of Table 3.3-4 for the'affected channel, and-TA = The value from Column TA (Total Allowance) of Table 3.3-4 for the affected channel.

c.

With an ESFAS instrumentation channel cr interlock inoperable, take the ACTION shown-in Table 3,3-3.

SEABROOK - UNIT 1 3/4 3-14

%5 POWER DISTRIBUTION LIMIT.;

BASES 3/4,2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and NUCLEAll ENTHALPY RISE HOT tiiEiniEI7ATT5R (Continued)

F"H will be maintained within its limits provided Conditions a, through S

d. above are mairtained.

The relaxation of F as 6 function of THERMAL POWER g

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

Fuel rod bowing reduces the value of DNBR.

Credit is available to offset this reduction in the generic mar 0in.

The generic margins, totaling 9.1% ONBR completely offset any rod bow penalties.

This msrgin includes the fnllowing:

a.

Design limit DNBR of 1.30 vs. 1.28, b.

Grid spacing (K,) of 0.046 vs. 0.059, c.

Thermal diffusion coefficient of 0.038 vs. 0.059, d.

DNBR r.ultiplier of 0.86 vs. 0.88, and e.

Pitch reduction.

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

When an F measurement is taken, an allowance for both experimental error g

and manufacturing tolerance must be made.

An allowance of 5% is appropriate for a full-core map taken with the Incore Detector Flux Mapping System, and a 3% allowance is appropriate for manufacturing tolerance.

The Radial _ Peaking Factor, F,y(Z), is m esured periodically to provide assurance that the Hot Channel Factor, F (Z), remains within its limit.

The F,y limit for RATED THERMAL POWER (F P)q as provided in the CORE OPERATING LIMITS REPORT per Specification 6.8.1.6 was determined from expected power control maneuvers over the full range of burnup conditions in the core.

When RCS Fh is measured, no additional allowances are necessary prior to comparisonwiththeestablishedlimitoiimeasurementerrorof4%forFhhas been allowed for in deterinination of the design DNBR value.

3/4.2.4 QUADMNT POWER TILT RATIO The purpose _of this specification is to detect gross changes 'n core power distribution between monthly incore flux maps.

During normal operation the QUADRANT POWER TILT RATIO is set equal to zero once acceptability of core l

peaking factors has been established by review of incore maps.

The limit of 1.02 is establisheo as an indication that the power distribution has changed enough to warrant further investigation.

L SEABROOK - UNIT ?.

B 3/4 2-3 Amendment No. g. 12 L

FEB18id

POWER DISTRIBUTION LlHITS BASES 3/4.2.5 DNB PARAMETERS The limits on the DNB-related parameters assure that each of the parameters is maintained within the normal steady-state envelope of operation assumed in the transient and accident analyses.

The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR of 1.30 throughout each analyzed transient.

Operating procedures include allowances for measurement and indication uncertainty so that the limits of 594.3 F for T and 2?05 psig for pressurizer are not exceeded.

avg The measurement error of 2.4% for RCS total flow rate is based upon per-forming a precision heat balance aad m ing the result to normalize the RCS flow rate indicators.

Potential fouling of the feedwater venturi which might not be detected could bias the result from the precision heat balance in a noncon-servative manner.

Therefore, a penalty of 0.1% for undetected fouling of the feedwater venturl is applied.

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

If detected, action shall be taken before performing subsequent precision heat balance measurements, i.e., either the effect of the fouling shall be quantified and compensated for in the RCS flow rate measurement or the venturi shall be cleaned to eliminate the fouling.

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 transic1t operation.

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

i SEABROOK - UNIT 1 B 3/4 2-4 Amendment No. 7, 12

, -, - - -.

ML20099G298

Text

Attachment:

Navigation menu

Search