ML20195D404

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Proposed Tech Specs,Modifying Sections 3.3.1.1 & 3.3.2.1 by Restricting Time That Reactor Protection or ESF Actuation Channel Can Be in Bypass Position to 48 H,From Indefinite Period of Time
ML20195D404
Person / Time
Site: Millstone Dominion icon.png
Issue date: 11/10/1998
From:
NORTHEAST NUCLEAR ENERGY CO.
To:
Shared Package
ML20195D402 List:
References
NUDOCS 9811180033
Download: ML20195D404 (50)


Text

.

Docket No. 50-336 B17492 o

i l

I Millstone Nuclear Power Station, Unit No. 2 Proposed Revision to Technical Specifications Reactor Protective and Engineered Safety Feature Actuation System Instrumentation j

Marked Up Pages l

I l

November 1998 l-t-

-9811190033 981110 P

l PDR ADOCK 05000336 4

p PDR

1 TABLE 2.2-1 35 EEACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS 2l-4 k

FUNCTIONAL UNIT IRIP SETPOINI ALLOWABLE VALUES 1.

Manual Reactor Trip Not Applicable Not Applicable e

E Power Level-High 2.

q four Reactor Coolant Pumps 5 9.6% above THERMAL POWER, s 9.7% Above THERMAL POWER, with a minimum setpoint of with a minimum of 514.7%.

n Operating of6 /06.4% of

~< 14.6% of RATED THERMAL

'of RATED THERMAL POWER, and a 2 aa d a '"* *" "

maximum of 5106.7% of POWER.

Smo mu.*a RATED THERMAL POWER.

t rown J

3.

Reactor Coolant flow -

2 91.7% of reactor coolant.

2 90.9% of reactor coolant flow flow with 4 pumps operating *.

with 4 pumps operating.

Low (1) 4.

Reactor Coolant Pump 1 830 rpm 2 823 rps Speed - Low (

[

5.

Pressurizer Pressure - High 5 2400 psia s.2408 psia 6.

Containment Pressure - High 5 4.75 psig 5 5.24 psig 7-S***" "'"' "r' ""r" 2 68 P

2 '72 '

0F y,

Low (2)

(5) i Ok 8.

Steam Generator Water 1 36.0% Water Level - each 2 35.2% Water Level - each yg Level - Low (5) steam generator steam generator l

Mg 9.

Local Power Density -

Trip setpoint adjusted to not Trip setpoint adjusted to

?*

High (3) exceed the limit lines of not exceed the limit lines l

Figures 2.2-1 and 2.2-2 (4).

of Figures 2.2-1 and s@

2.'2-2 (4).

4&

w y

g

  • Design Reactor Coolant flow with 4 pumps operating is the lesser of either:

o

a. The reactor coolant flow rate measured per Specification 4.E.6.1, or f

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b. The minimum value specified in the CORE OPERATING LIMITS REPORT.

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q

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25 3 r-i

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TABLE 2.2-1 E

REACTOR PROTECTIVE INSTRUNENTATION TRIP SETPOINT LINITS j

g FUNCTIONAL UNIT TRIP SETPOIPE ALLOWABLE VALUES t

10. Thermal Margin / Low Pressure (1)

Four Reactor Coolant Pumps Trip setpoint adjusted to not Trip setpoint adjusted Operating exceed the limit lines of to not exceed the ifeft Figur's 2.2-3 and 2.2-4 (4)..

lines of Figures 2.2-3 i

e and 2.2-4 (4).

II.

Loss of Turbine--Hydraulic l 500 psig 2 500 psig Fluid (3) Pressure - Low TABLE NOTAILOH (1) Trip may be bypassed below 5% of RATED THERMAL POWER; bypass shall be automatically removed w is 2 5% of RATED THERMAL POWER.

(2) Trip may be manually bypassed below 780 psia when all CEAs are fully inserted; bypass shall be automatically l

removed at or above 780 psia.

(3) Trip may be bypassed below 15% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER

't IS 215% of RATED THERMAL POWER.

(4) Calculations of the trip setpoint includes measurements, calculational and processor uncertainties, and dynamic g.

allowances.

c i

k (5) Each of four channels actuate on the auctioneered output of two transmitters, one from each steram generator. E E

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June 10,1996

.A/0 C H&vGE 3/4.3 INSTRUMENTATION

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po e revrogenerzw 3/4.3.1 REACTOR PROTECTIVE INSTRUMENTATION ON LIMIT 1HG CONDITION FOR OPERATION 3.3.1.1 As a minimum, the reactor protective instr'unentation channels and I

bypasses of Table 3.3-1 shall be OPERABLE.

I APPLICABILITY: As shown in Table 3.3-1.

l ACTION:

As shown in Table 3.3-1.

SURVEILLANCE REQUIREMENTS

\\

4.3.1.1.1 Each reactor protective instrumentation ch'annel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the modes and at the freq'uencies shown in Table 4.3-1.

4.3.1.1.2 The logic for the bypasses shall be demonstrated OPERABLE during the at power CHANNEL FUNCTIONAL TEST of channels affected by bypass

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operation. The total bypass function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by bypass operation.

4.3.1.1.3 The REACTbR TRIP SYSTEM RESPONSE TIME of each reactor trip function shall be demonstrated to.be within its limit at least once per 18 months.

Heutron detectors are exempt from response time testing.

Each test l

shall include at least one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific reactor trip function as shown in the " Total No. of Channels" column of Table 3.3-1.

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MILLSTONE - UNIT 2 3/4 3-1 Amendment No. 77.198 0246

TADLE 3.3-1 REACTOR PROTECTIVE INSTRUMENTATION x-HINIMilM G

TOTAL NO.

CHANNELS CllANNELS APPLICADLE g

M FUNCTIONAL UNIT _

0F CHANNELS TO TRIP OPERABLE MOI 1CS ACTIOft 1.

Manual Reactor Trip 2

1 2

1, 2 and +

1 h.

f 2.

Power Level - High,

4 2 (f) 3

1. 2. 3(d) 2 3.

Reactor Coolant flow - Low 4

2(a) 3

1. 2 (c) 2 4.

Pressurizer Pressure - liigh 4

2 3

1, 2 2

r 5.

Containment Pressure - lifgh 4

2 3

1. 2 2

6.

Steam Generator Pressure - Low 4 2(b) 3 1, 2 2

l y

7.

Steam Generator Water Level - Low 4

2 3

1, 2 2

8.

Local Power Density - High

'4 2(c) 3 1

2 9.

Thermal Margin / Low Pressure 4

2(a) 3

1. 2 (c) 2

[

10. Loss of Turbine--Ilydraulic Fluid Pressure - Low 4

2(c) 3 1

?

n g

4 i

%l s

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TABLE 3.3-1 (Continued)_

REACTOR PROTECTIVE INSTRUMENTATION x

r"-

G MINIMUM TOTAL NO.

CHANNELS CHANNELS APPLICABLE.

MODES ACTION M

0F CHANNELS _

TO TRIP OPERABLE _ _

FUNCTIONAL UNIT _

i c:

11. Wide Range Logarithmic Neutron Flux Monitor - Shutdown 4

0 2

3,4,5

'4 4

2(a) 3 1,'2(e)-

2 y

12. Underspeed - Reactor Coolant Pumps e

w 9

R ta)b ows t c Q^Q F

a O

[

H 5

n h%

3 G

f

?

m S

G U

t S %

t E

4md M,

'^G l

TABLE 3.3-1 (Continued)

I TABLE NOTATION

[

  • With the' protective system trip breakers in the closed position and the CEA drive system capable of CEA withdrawal.

[

(a) Trip may be bypassed below 5% of RATED THERMAL POWER; bypass shall be i

automatically removed when THERMAL POWER is 15% of RATED THERMAL POWER.

(b) Trip may be manually bypassed below 780 psia when all CEAs are fully inserted; bypass shall be automatically removed at or above 780 psia.

{

(c) Trip may be bypassed below 15% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is 115% of RATED THERMAL POWER.

i r

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(d) Trip does not need to be operable if all the control rod drive mechanisms i

are de-energized or if the RCS boron concentration is greater than or i

equal to the refueling concentration of Specification 3.9.1.

(e)

Trip may be bypassed during testing pursuant to Special Test Exception 3.10.3.

(f) AT Power input to trip may be bypassed below 5% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is 15% of RATED l.

THERMAL POWER.

ACTION STATEMENTS l

l ACTION 1 -

With the number of channels OPERABLE one less than requ' ired by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in HOT STANDBY within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and/or open the protective system trip breakers.

ACTION 2 -

With the number of OPERABLE channels one less than the Total Number of Channels -:nd with the Tl E"#AL P^WER level:

i 15% er "ATED TMEP. MAL P0"EP., Medi:tely phcc the inoper-abh-chrael " the typ;ss;d ccaditien; restere the inupereble ch nnel te GFEisABLE si.ai.us privi te increasing-THERMAL-POWER---

bev: 5% ef PATED 'ME"?AL POWER.

4.

l' i 5% :f PATED vugpuft poi;ER, operation may continue with th:

'neper:bl: ch;nn:1 in th; typ::::d :;nditica, provided the following conditions are satisfied:

.TA) SER r f) i I

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'Mll.tSTONE - UNIT 2 3/4 3-4 AmendmentNo.OM,7/,JJO[h L..

INSERT A - Paae 3/4 3-4 a.

The inoperable channel is placed in either the bypassed or tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The inoperable channel shall either be restored to OPERABLE status, or placed in the tripped condition, within 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />, b.

Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, all functional units receiving an input from the inoperable channel are also declared inoperable, and the appropriate actions are taken for the affected functional units.

c.

The Minimum Channels OPERABLE requirement is met; however, one additional channel may be removed from service for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, provided one of the inoperable channels is placed in the tripped condition.

i 4

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- Auyuai. 1,1075 TABLE 3.3-1 (Continued)

-l.)

ACTION STATEMENTS 1.

All Inctional unite eceiving an in t from t

bypassed chan are also pla in the ypassed condi n.

The Minim

_hannels OPERAB requirement is met; how er, one additio channel may be remove from service for p to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> fo surv lance testing p Specification

.3.1.1 pr ided one of the

  • operable channe is aced in the trip d condition.

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ACTION 3

'With the numb of OPERABLE ch nels one less tha the Total Numb of Channels and ith the THERMAL P ER level:

l a.

5% of RATED THE L POWER, innedia ly place the x/ 07~

Inoperable cha 1 in the bypassed ondition, restore gggg the inoperab channel.to OPERAB status prior to increasing ERMAL POWER above - of RATED THERMAL i

POWER.

)

b.

> 5% f RATED THERMAL POW, power operation cop inue.

~

ACTION 4 With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, imniediately verify compliance with the SHUTDOWN MARGIN requirements of Specification 3.1.1.1 or 3.1.1.2, as applicable, l

and at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> thereafter.

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MILLSTONE - UNIT 2 3/4 3-5

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v TABLE 4.3-1 REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE RE EIREMENTS M

CllANNEL MODES IN WillCll

O5 CIIANNEL CilANNEL FUNCTIONAL SURVEiLl.ANCE CHECK CALIBRATION TEST __

REQ (!!PID FUNCTIONAL UNIT g

N.A.

N.A.

S/U(1)

N.A.

[

1.

Manual Reactor Trip 2.

Power Level - liigh D(2),M(3),Q H

1, 2, 3*

l i

S i

a.

Nuclear Power S

D(4),O M

b.

AT Power 1, 2

[5)

S R

H 3.

Reactor Coolant Flow - Low 4.

Pressurizer Pressure - liigh 5

R M

I, 2 5.

Containment Pressure - liigh 5

R H

1, 2 S

R H

I, 2

[

6.

Steam Generator Pressure - Low S

R M

1, 2 7.

Steam Generator Water level - Low g

S R

H I

8.

Local Power Density - High 5

R H

1, 2 N

A 9.

Thermal Margin / Low Pressure 2

P

10. Loss of Turbine--Hydrauilc N.A.

4:A-5/ll(l)

N.A.

h Fluid Pressure - Low

(

c 1

(

l

TABLE 4.3-1 (Continued) -

REACTOR PROTECTIVE INSTRUMENTATION' SURVEILLANCE REQUIREMENTS

-FU CllANNEL MODES IN WillCil E-CilANNEL Cl!ANNEL FUNCTIONAL SURVEILLANCE

"' FUNCTIONAL UNI:1 CilECK CALIBRATION TEST REQUIRLU E 11. Wide Range Logarithmic Neutron S

N.A.

S/U(1)-

.3, 4, 5 and

  • 4 Flux Monitor R(s)

~

12. Underspeed - Reactor S

R M

1, 2 Coolant Pumps

13. Reactor Protection System Logic N.A.

N.A.

M and S/U(1) 1, 2

14. Reactor Trip Breakers N.A.

N.A.

M 1, 2 and

  • M.

OJ k

-:e a

a it

?

.I ii

[

Scptemt,er 2, 570-

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TABLE 4.3-1 (Continued) i TABLE NOTATION With reactor trip breaker closed.

(1)

If not performed in previous 7 days.

(2)

Heat balance only, above 15% of RATED THERMAL POWER; adjust "Huclear Power Calibrate" potentiometers to make 8

nuclear power signals agree with calorimetric calculation.

During PHYSICS TESTS, these daily calibrations of nuclear 14 power and AT power may be suspended provided these calibra-tions are performed upon reaching'each major test power plateau and prior to proceeding to the next major test power plateau.

(3)

Above 15% of RATED THERMAL POWER, recalibrate the excore detectors which monitor the AXIAL SHAPE INDEX by using 8

~

the incore detectors or restrict THERMAL POWER during subsequent operations to < 90% of the maximum allowed l

THERMAL POWER level with the existing Reactor Coolant Pump combination.

-)

(4) -

Above 15% of RATED THERMAL POWER, adjust " AT Pwr Calibrate" potentiometers to null " Nuclear Pwr - AT Pwr". :During PHYSCIS TESTS, these daily calibrations of nuclear power and 14 AT power may be suspended provided these calibrations are performed upon reaching each major test power plateau and prior to proceeding to the next major test power plateau.

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MILLSTONE - UNIT 2 3/4 3-9

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June 10,1996 JNSTRUNENTATION d'0 C /hW6E Fr>g.Tsrcan drz w 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATIONoxy

[

i LINITING CONDITION FOR OPERATION 3.3.2.1 The engineered safety feature actuation system instrumentation channels and bypasses shown in Table 3.3-3 shall be OPERABLE with their

~

trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4.

l APPLICABILITY: As shown in Table 3.3-3.

ACTION:

With an engineered safety feature actuation system instru-a.

i mentation channel trip setpoint less conservative Wn the value shown in the Allowable Values column of Tabic 2.3-4, either adjust the trip setpoint to be consistent with the value specified in the Trip Setpoint column of Table 3.3-4 within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or declare the chantiel inoperable and take the ACTION 4

shown in Table 3.3-3.

1 b.

With an engineered safety feature actuation system instru-mentation channel inoperable, take the ACTION shown in Table 3.3-3.

(

SURVEILLANCE REQUIREMENTS 4.3.2.1.1 Each engineered safety feature acutation system instrumen-tation channel shall be demonstrated OPERABLE by the perfomance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the modes and at the frequencies shown in Table 4.3-2.

4.3.2.1.2 The logic for the bypasses shall be demonstrated OPERABLE during the at power CHANNEL FUNCTIONAL TEST of channels affects by bypass operation. The total bypass function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by bypass operation.

MILLSTONE - UNIT 2 3/4 3-10 Amendment No.198 0247

4 March 1,1979 A/oct/4NGE INSTRUMENTATION

      1. D
  • g(y SURVEILLANCE REQUIREMENTS (Continued) 4.3.2.1.3 The ENGINEERED SAFETY FEATURES RESPONSE TIME of'ench ESF function'shall be demonstrated to be within the limit at least once per 18 months. Each test shall include at least one channel per function such that all channels are tested at.least once every N times 18. months where H is the total number of redundant channels in a specific ESF function as shown in the " Total,No.,of Chanpels". Column of Table 3.3-3.

l 4.3.2.1.4 The trip value shall be,such that the containment purge effluent shall not result in calculated concentrations of radioactivity offsite in excess of 10 CFR Part 20, Appendix B. Table 116 purposes of calculating this trip value, a x/Q = 5.8 x 10' sec/mForghe shall beused,whenthesystgmisagignedtopurgethroughthebuildingvent and a X/Q = 7.5 x 10~ sec/m shall be used when the system is aligned to purge through the Unit 1 stack, the gaseous and aprticulate (Half Lives greater than 8 days) radioactivity shall be,asusmed to be Xe-133 and Cs-137,5respectiv.ely. However, the setpoints shall be no greater than 5 x 10 cpm.

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MILLSTONE - UNIT 2 3/4 3'11

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I February 22, 1982 s,

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TABLE 3.3-3 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRLHENTATION_

.1F G

MINIMUM CHANNELS APPLICABLE h

TOTAL NO.

CHANNELS OPERABLE _

MODES ACTION OF CHANNELS _

TO TRIP 1 m

i FUNCTIONAL UNIT 1

l 61

.g SAFETY. INJECTION (SIAS)

Manual (Trip Buttons) 2 1

2 1,2,3,4 p

1.

a.

y 2

l 61 4

2 3

1' 2' 3 b.

Containment Pressure -

High 1

3N 2

l 61 c.

Pressurizer Pressure -

Manual (Trip Buttons) 2 1

2-1,2,3,4 1

l 61 2.

CONTAINMENT SPRAY (CSAS) a.

w r

b.

Containment Pressure --

2(b) 3 1,2,3 2

l. 61 2

4

.High - High w

m CONTAINMENT ISOLATION (CIAS) 2 1

2 1, 2, 3, 4 1

[ 61 3.

a.

Ma C 5 (Trip

~

b.

Manual SIAS (Trip 1-2 1,2,3,4 1

J. 61 l

2

, Buttons) l 61 l

2 0 ) g 1, 2, 3 c.

Centainment Pressure -

2 3

r fa 4

High,

o Tq n d.

Pressurizer Pressure -

1,2(e),3(a) 2 l 61

~

4 2

3 Low g

y t

^

.m

v TABLE.3.3-3 (Continued) 2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM 83 BF TOTAL NO.

CNANNELS CilANNELS APPLICABLE OF CllANNELS TO TRIP.

OPERABLE MODES ACTION-h FUNCTIONAL UNIT x

4.

MAIN STEAM LINE ISOLATION E

Manual MSI (Trip Buttons) 2 1

2 I, 2, 3, 4 I

Z a.

b.

Containment Pressure -

4 2

3 I, 2, 3 2

liigh 4

2 3

1,2,3(c) 2 c.

Steam Generator Pressure - Low S.

ENCLOSURE BUILDING FILTRATION (EBFAS)

Manual EBFAS (Trip 2

1 2

I, 2, 3, 4 I

a.

Buttons) y b.

Manual SIAS (Trip 2

1 2

I, 2, 3, 4 1

Buttons) 4 2

3 I, 2, 3 2

c.

Containment Pressure-liigh o)4 4

2 3

1,2,3(a) 2 4 o

d.

Pressurizer Pressure-Low a

g l

6.

CONTAINMENT SUMP RECIRCULATION (SRAS) 1Nh

,S 3 m b h

J

^

2 I

.2 I, 2, 3, 4-I Manual SRAS (Trip e

a.

Buttons) d j

q w

l 4

2 3

I, 2, 3 4

b.

Refueling Water Storage E

Tank - Low

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TABLE 3.3-3 (Continued)

Q3

.h.

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION

.' E MINIMUM m

TOTAL.NO.

CHANNELS CHANNELS APPLICABLE i

g 0F CHANNELS TO TRIR OPERABLE MODES ACTION FUNCTIONAL UNII q

7.

CONTAINMENT PURGE VALVE ISOLATION Containment Radiation-5, 6 a.

3 High Gaseous Monitor 2

1 1

3 Particulate Monitor 2

1 1

8.

LOSS OF POWER s

4.16 kV Emergency Bus

^

a.

Undervoltage (Under-ma voltage relays) -

level one 4/ bus 2/ Bus 3/ bus I, 2, 3 2

a b.

4.10 kV Emergency Bus Undervoltage (Under-k level two 4/ Bus 2/ Bus 3/ Bus 1, 2, 3 2

voltage relays) -

SD a

(

kn R

=>

o e

a 2?

0 D

~

cn

~

?R y

4i E

m G

f

TABLE 3.3-3 (Continued) gz C

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUNENTATION U

i MINIMUM

-TOTAL NO.

CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION Y).

9.

AUXILIARY FEEDWATER n

a.

Manual 1/ pump 1/ pump 1/ pump I, 2, 3 1

b.

Steam Generator 4

2 3

1, 2, 3 2

l Level - Low R

i 0;

k k

b en n 4

m 33 e

3-4 S

i I

w$

5 3

?

^

i w

D 8

a R

S b

4, n

..--.g--...-.--..--_-

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

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

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

.~.IUUU vaisunny it TABLE 3.3-3 (Continued)

TABLE NOTATION

{

(a) Trip function may be bypassed when pressurizer pressure is < 1750 pstat bypass shall be automatically removed when pressurizer pressure is 11750 psia.

(b) An SIAS signal is first necessary to enable CSAS logic.

(c) Trip function may be bypassed below 600 psia; bypass shall be automatically removed at or above E00 psia.

(d) Deleted

}

(e) Trip may be bypassed during testing pursuant to Special Test Exception 3.10.3.

ACTION STATEMENTS l

ACTION 1 -

With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in COLD SHUTDOWN within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

ACTION 2 -

With the number of OPERABLE channels one less than the Total Humber of Channels :nd with the pre::urf::E pretture:

(

< 1700 psie; i=;diEtely p1;c; th: inoperable dannel in the byp::: d condititm;, resture the inoptrable-chan6ebte

-OPERABI statc; prier te increising the pr;.::urizer

-pruswe :beve 1750 p:f:.

4.

i 1700 psia, operation may continue with th: in:;;r: bis ch:nr.:1 in th: byp::::d ::r.ditica, provided the following conditions are. satisfied:

i 4

1.

All feLitier.;l unit: M :iving tr. input fr;; th:

byp::::d chtanal are else pieced in the 1,jpstsed-conditt0.

1

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-2.

The liini;;; Cher.neli CEEEAELE re;isir;;;at is set;

)

i h:u:ver, ::: :dditi:::1 h:.nn:1 ::y be reseved fre;

Orlic: fer c p;r Sp::iMe:p to 2 hessi foi ini wwillsise i,sstiLi tien 4.3.t a previded ene ;f the

-inep:rable thittel: 1: pitted is the tri;;:d

-cenditi::.

4 I/>SEE r 93

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MILLSTONE - UNIT 2 3/4 3-16 AmendmentNo.JU,777.k

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l INSERT B - Paae 3/4 3-16 l

l l

a.

' The inoperable channel is placed in either the bypassed or tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The inoperable ' channel shall either be restored to OPERABLE status, or placed in the tripped condition, within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, l

b.

Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, all functional units receiving an input from the inoperable channel y

are also declared inoperable, and the appropriate actions are taken for the i

affected functional units.

l-c.

.The Minimum Channels OPERABLE requirement is met; however, one additional channel may be removed from service for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, provided I.

one of the inoperable channels is placed in the tripped condition.

l l

l E

i i

i

Tarwari 17,133G -

IABLE 3.3-3 fContinued)

)

ACTION 3 -

With less than the minimum channels OPERABLE the containment purge valves are to be maintained closed.

ACTION 4 -

With the number of OPERABLE channels one less than the Total Humber of Channels and with the pressurizer pressure:

3 i

a.

< 1750 psia: immediately place the inoperable channel in the bypassed condition; restore the inope.able channel to OPERAB.E status prior to increasing the pressurizer pressure above 1750 psia.

b.

11750 psia, operation may continue with the inoperable channel in the bypassed condition, provided the following condition is satisfied:

1.

The Minimum Channels OPERABLE requirement is met; however, one additional channel may be removed from service for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for survelliance testing per Specification 4.0.0.1-provided M of the inoperable channelsfare placed in the bypassed condition.

(

7

.)

y,3,3,lp i

I l

l l

i

)

{1glSTONE-UNIT 2 3/4 3-17 AmendmentNo.JU.J77.[M l

v U

w-3 TABLE 4.3-6 P

REMOTE SHUTDOWN MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL i

7 CHECK CALIBRATION _

INSTRUMENT _

4 M

c t

a 1.

Wide Range Logarithmic Neutron Flux N

2.

Reactor Trip Breaker Indication M

N.A.

M R

3.

Reactor Cold Leg Temperature 4.

Pressurizer Pressure

~

M R

a.

Low Range M

M R

b.

High Range i

w M

R 5.

Pressurizer Level M~

R 6.

Steam Generator Level M

R 7.

Steam Generator Pressure i

t-xc luch d bm The f/HNNEL C ALZ $ffyTZw.

{

k A/cufrou c}r/re/oq air f

.fa 21

-durie 10, isso 3/4.3 INSTRUENTATION BASES

}Zi.t3.1 AND 3/4.3.2 PROTECTIVE AND ENGINEERED SAFETY FEATURES 1ESF) INSTRUMENTATION The OPERABILITY of the protective and ESF instrumentation systems and bypasses ensure that 1) the associated ESF action and/or reactor trip will be initiated when the parameter monitored by each channel or combination thereof exceeds its setpoint, 2) the specified coincidence logic is maintained,

3) sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance, and 4) sufficient system functional i

capability is available for protective and ESF purposes from diverse parameters.

The OPERABILITY of these systems is required to provide the overall reliability, redundance and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions.

integrated operation of each of these systems is consistent with the The l

assumptions used in the accident analyses.

f fA/SfRr C]

i The surveillance requirements specified for these systems ensure that the overall system functional capability is maintained comparable to the original design standards.

l frequencies are The: periodic surveillance tests performed at the minimum emonstrate this capability.

The measurement of response time at the specified frequencies provides assurance that the protective and ESF action function associated with each channel is completed eithin the time limit assumed in the accident analyses.

No credit was taken in the analyses for those channels with response times indicated as not applicable.

Feature response times are contained in the Millstone Unit No. 2 Techn Requirements Manual.

10CFR50.59 review as well as a review by the Plant Operations Review Committee.

The containment airborne radioactivity monitors (gaseous and particulate) i detection of high radioactivity levels in the containment.are provided to Closure of these valves prevents excessive amounts of radioactivity from being released to the l

environs in the event of an accident. p i

i The a <k

/09.c h tLs A. Ac is 1. od of % 19ck SMc~s/ 3 of Tabt< 3, 3 - 3 aOm es /"y"d'r fu % med pyx e Ausa4 1

(

l MILLSTONE - UNIT 2 B 3/4 3-1 Amendment No. W. M. //g 0261 l

INSERT C - Pace B 3/4 3-1 I

Action Statement 2 of Tables 3.3-1 and 3.3-3 requires an inoperable Reactor Protection System (RPS) or Engineered Safety Feature Actuation System (ESFAS) channel to be placed in the bypassed or tripped condition within i hour.

The inoperable channel may remain in the bypassed condition for a maximt'm of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

While in the bypassed condition, the affected functional unit trip coincidence will be 2 out of 3. After 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the channel must either be declared OPERABLE, or placed in the tripped condition. If the channel is placed in the tripped condition, the affected functional unit trip coincidence will become 1 out of 3. One additional channel may be removed from service for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, provided one of the inoperable channels is placed in the tripped condition.

Plant operation with an inoperable pressurizer high pressure reactor protection channel in the tripped condition is restricted because of the potential inadvertent opening of both pressurizer power operated relief valves (PORVs) if a second pressurizer high pressure reactor protection channel failed while the first channel was in the tripped condition. This plant operating restriction is contained in the Technical Requirements Manual.

l i

i l

l 1

~

INSERT D - Pace B 3/4 3-1 The surveillance testing verifies OPERABILITY of the RPS by overlap testing of the four interconnected modules: measurement channels, bistable trip units, RPS logic, and reactor trip circuit breakers. When testing the measurement channels or bistable trip units that provide an automatic reactor trip function, the associated RPS channel will be removed from service, declared inoperable, and Action Statement 2 of Technical Specification 3.3.1.1 entered.

When testing the RPS logic (matrix testing), the individual RPS channels will not be affected. Each parameter within each RPS channel supplies three contacts to make up the 6 different logic ladders / matrices (AB, AC, AD, BC, BD, and CD). During matrix testing only one logic matrix is tested at a time. Since j

oach RPS channel supplies 3 different logic ladders, testing one ladder matrix at a time will not remove an RPS channel from the overall logic matrix. Therefore, matrix testing will not remove an RPS channel from service or make the RPS channel inoperable. It is not necessary to enter an action statement while performing matrix testing. This also applies when testing the reactor trip circuit breakers since this test will not remove an i

RPS channel from service or make the RPS channel inoperable.

i 1

i E

1

l 10/7/94 INSTRUMENTATION

\\

R>R NFM M n w BASES o.v t. P 3/4.3.1 AND 3/4,3.2 PROTUAlVE isia ENGINEERED SAFETY FEATURES (ESF)

INSTRUMENTATION (Continued)

The maximum allowable trip value for these monitors corresponds to calculated concentrations at the site bounoary which would not exceed the concentrations listed in 10 CFR Part 20, Appendix B, Table II. Exposure for a year to the concentrations in 10 CFR Part 20, Appendix B.

Table corresponds to a total body dose to an individual of 500 mram which is well below the guidelines of 10 CFR Part 100 for an individual at any point on the exclusion area boundary for two hours.

Determination of the monitor's trip value in counts per minute, which is the actual instrument response, involves several factors including:

1) the atmospheric dispersion (x/Q), 2) isotcpic composition of the sample, 3) sample J

flow rate, 4) sample collection efficiency, 5) counting efficiency,dand 6) the background radiation level at the detector.

The x/Q of 5.8 x lo sec/m is the highest annual average x/Q estimated for the site boundary (0.48 miles in j

the NE sector) for vent releases from the containment. and 7.5 x 10 sec/m' is

)

4 the highest annual average x/Q estimated for an off-site location (3 miles in

~

the NNE sector) for releases from the Unit I stack.

This calculation also assumes that the isotopic composition is xenon-133 for gaseous radioactivity

)-

and cesium-137 for particulate radioactivity (Half Lives greater than 8 days).

J The' upper limit of 5 x 10' cpm is approximately 90 percent of full instrument 1

scale.

$MS Loaic Modification i

Action Statement ~4 of Table 3.3-3, which applies only to-the SRAS logic,. '

i specifies that during surveillance testing the second inoperable channel must also be placed in the bypassed condition. For the SRAS logic, placing the second inoperable channel in the tripped condition (as in Action Statement 2) could result in the false generation of. a SRAS signal due to an addi.tional failure which causes 'a trip signal in either of the remaining channels at the onset of a LOCA.

The false generation.of the SRAS signal leads to unacceptable consequences for LOCA mitigation.

With Action Statement 4, during the two-hour period when two channels are bypassed, no additional failure can result in the false generation of the SRAS signal. However, an additional failure that prevents a trip of either of the two i

remaining channels may prevent the generation of a true SRAS signal while in this Action Statement.

If no SRAS is generated at the appropriate time, operating proceduresinstruct the operator to ensure that the SRAS actuation occurs when the refueling water storage tank level decreases.

Due to the limited period of i

vulnerab?1ity, and the existence of operator requirements to manually initiate i

an SRAS 1f an automatic initiation does not occur, this risk is considered g acceptabic.

)b MILLSTONE - UNIT 2 B 3/4 3-2 Amendment No. U7 179 0146 a

n

l Docket No. 50-336 817492 Millstone Nuclear Power Station, Unit No. 2 Proposed Revision to Technical Specifications Reactor Protective and Engineered Safety Feature Actuation System Instrumentation Retyped Pages November 1998 f-

a TABLE 2.2-1

$h REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS G

h FUNCTIONAL UNIT TRIP SETPOINT ALLOWA8LE VALUES 1.

Manual Reactor Trip Not Applicable Not Applicable E

2.

Power Level-High

[

Four Reactor Coolant Pumps s 9.6% above THERMAL POWER, s 9.7% Above THERMAL POWER, Operating with a minimum setpoint of with a minimum of s 14.7%

s 14.6% of RATED THERMAL of RATED THERMAL POWER, and a POWER, and a maximum of maximum of s 106.7% of s 106.6% of RATED THERMAL RATED THERMAL POWER.

POWER.

3.

Reactor Coolant Flow -

2 91.7% of reactor-coolet 2 90.9% of reactor coolant flow Low (1) flow with 4 pumps operating *.

with 4 pumps operating.

4.

Reactor Coolant Pump 2 830 rpm 2 823 rpm Speed - Low (1) l 1

5.

Pressurizer Pressure - High 1 2400 psia 5 2408 psia 6.

Containment Pressure - High 5 4.75 psig s 5.24 psig

$(

7.

Steam Generator Pressure -

2 680 psia 2 672 psia g

Low (2)

(5) 8.

Steam Generator Water 2 36.0% Water Level - each 2'35.2% Water Level - each Level - Low (5) steam generator steam generator k.E 9.

Local Power Density -

Trip setpoint adjusted to not Trip setpoint adjusted to High (3) exceed the limit lines of not exceed the limit lines y

QP Figures 2.2-1 and 2.2-2 (4).

of Figures 2.2-1 and P.

2.2-2 (4).

?

4 i

  • Design Reactor Coolant flow with 4. pumps operating is the lesser of either:

4

a. The reactor coolant flow rate measured per Specification 4.2.6.1, or
b. The minimum value specified in the CORE OPERATING LIMITS REPORT.

l

~

TABLE 3.)-1 p

~b REACTOR PROTECTIVE INSTRUMENTATION v

w E

D1 h.

MINIMUM TOTAL NO.

CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE-MODES ACTION

]

1.

Manual Reactor Trip 2

1 2

-1,.2 and

  • 1 2.

Power Level - High 4

2(f) 3 1,2,3(d) 2' 3.

Reactor Coolant Flow - Low 4

2(a) 3 1, 2 (e) 2 4.

Pressurizer P: essure - High

'4 2

3 1, 2 2

N*

5.

Containment Pressure - High 4

2 3

1, 2 2

A 6.

Steam Generator Pressure - Low 4

2(b) 3 1, 2 2

7.

Steam Generator Water Level - Low 4

-2 3

1, 2 2

8.

Local Power Density - High 4

2(c) 3 1

2 9.

Thermal Margin / Low Pressure 4

2(a) 3 1,2(e) 2 E

10.

Loss of Turbine - Hyraulic 5

Fluid Pressure - Low 4

2(c) 3 1

2 l

A

.E N

w

~

TABLE 3.3-1 (Continued)

TABLE NOTATION

  • With the protective system trip breakers in the closed position and the CEA drive system capable of CEA withdrawal.

(a) Trip may be bypassed below 5% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is 1 5% of RATFD THERMAL POWER.

(b) Trip may be manually bypassed below 780 psia when all CEAs are fully inserted; bypass shall be automatically removed at or above 780 psia.

(c) Trip may be bypassed below 15% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is 115% of RATED THERMAL POWER.

(d) Trip does not need to be operable if all the *ontrol rod drive mechanisms are de-energized or if the RCS boron concentration is greater than or equal to the refueling concentration of Specification 3.9.1.

(e) Trip may be bypassed during testing pursuant to Special Test Exception 3.10.3.

(f) -AT Power input to trip may be bypassed below 5% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is 15% of RATED THERMAL POWER.

ACTION STATEMENTS ACTION 1 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in HOT STANDBY within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and/or open the protective system trip breakers.

ACTION 2 - With the number of OPERABLE channels one less than the Total Number of Channels, operation may continue provided the following conditions are satisfied:

a.

The inoperable channel is placed in either the bypassed or tripped condition within I hour. The inoperable channel shall either be restored to OPERABLE status, or placed in the tripped condition, within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

b.

Within I hour, all. functional units receiving an input from the inoperable channel are also declared inoperable, and the appropriate actions are taken for the affected functional units.

c.

The Minimum Channels OPERABLE requirement is met; however, one additional channel may be removed from service for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, provided one of the inoperable channels is placed in the tripped condition.

MIgLSTONE-UNIT 2 3/4 3-4 Amendment No. 7, 77, 77, JJ7, J77,

TABLE 3.3-1 (Continued)

ACTION STATEMENTS l

i ACTION 3 -

NOT USED l

ACTION 4 -

With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, immediately verify compliance with the SHUTDOWN MARGIN requirements of Specification 3.1.1.1 or 3.1.1.2, as applicable, and at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> thereafter.

l l

l l

i NILLSTONE - UNIT 2 3/4 3-5 Amendment No.

0302

1 t

TABLE 4.3-1 r-U' REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE RE0VIREMENTS E

~

CHANNEL MODES IN WHICH-CHANNEL CHANNEL FUNCTIONAL ~

SURVEILLANCE-E FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED A

S/U(1)

N.A.

m.

1.

Manual Reactor Trip N.A.

N.A.

2.

Power Level - High

a. Nuclear Power S

D(2),M(3),Q(5)

M 1, 2, 3*

l

b. AT Power S

D(4),Q M

1 3.

Reactor Coolant Flow - Low S

R M

1, 2 1-,2 4.

Pressurizer Pressure - High S

R.

M 5.

Containment Pressure - High S

R M

1, 2 i

6.

Steam Generator Pressure - Low S

R M

1,-2 7.

Steam Generator Water S

R M

1, 2 Level - Low 8.

Local Power Density - High S

R M

1 9.

Thermal Margin / Low Pressure S

R M

1, 2 1

10. Loss of Turbine--Hydraulic

{

Fluid Pressure - Low N.A.

R S/U(1)

N.A.

.l 5

5 a

F.

,$p TABLE 4.3-1 (Continued)

U REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE REQUIREMENTS E

m CHANNEL MODES IN WHICH E-CHANNEL-CHANNEL FUNCTIONAL SURVEILLANCE A

FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED m

11. Wide Range Logarithmic Neutron S

R(5)

S/U(1) 3, 4, 5 and

  • l Flux Monitor
12. Underspeed - Reactor S.

R M

1, ' 2 Coolant Pumps

13. Reactor Protection System Logic N.A.

N.A.

M and S/U(1) 1, 2 w

14. Reactor Trip Breakers N.A.

N.A.

M 1, 2 and

  • w E

Ea F.

m

..m m

m

_____;___m____,-._____.r

,-w

i TABLE 4.3-1 (Continegdl TABLE NOTATION

- With reactor trip breaker closed.

(1)

If not performed in previous 7 days.

f (2)

Heat balance only, above 15% of RATED THERMAL POWER;-

adjest " Nuclear Power Calibrate" potentiometers to make nuclear power signals agree with calorimetric calculation.

During PHYSICS TESTS, these daily calibrations of nuclear power and AT power may be suspended provided these Calibra-tions are performed upon reaching each major test power plateau and prior to proceeding to the next major test power plateau.

(3) - Above 15% of RATED THERMAL POWER, recalibrate the excore detectors which monitor the AXIAL SHAPE INDEX by using the incore detectors or restrict THERMAL POWER during subsequent operations to s 90% of the maximum allowed THERMAL POWER level with the existing Reactor Coolant Pump combination.

(4) - Above 15% of RATED THERMAL POWER, adjust "AT Pwr Calibrate" potentiometers to null " Nuclear Pwr - AT Pwr".

During PHYSICS TESTS, these daily calibrations of nuclear power and AI power may be suspended provided these calibrations are performed upcn reaching each major test power plateau and prior to proceeding to the next major test power plateau.

(5)

Neutron detectors are excluded from the CHANNEL CALIBRATION.

l h

MILLSTONE - UNIT 2 3/4 3-9 Amendment No. 7. JJ, 0302

T TABLE 3.3-3 (Continued)

TABLE NOTATION (a) Trip function may be bypassed when pressurizer pressure is < 1750 psia; bypass shall be automatically removed when pressurizer pressure is 21750 psia.

(b) An SIAS signal is first necessary to enable CSAS logic.

(c) Trip function may be' bypassed below 600 psia; bypass shall be automatically removed at or above 600 psia.

(d) Deleted (e). Trip may be bypassed during testing pursuant to Special Test Exception i

3.10.3.

ACTION STATEMENTS ACTION 1 -

With the number of OPERABLE channels one less than the Total l

Number of Channels, restore the inoperable channel to OPERABLE j

status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in COLD SHUTDOWN within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, i

ACTION 2 -

With the number of OPERABLE channels one less than the Total Number of Channels, operation may continue provided the following conditions are satisfied:

a.

The inoperable channel is placed in either the bypassed or tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The inoperable channel shall either be restored to OPERABLE status, or placed in the tripped condition, within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

.b.

Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, all functional units receiving an input from the inoperable channel are also declared inoperable, and the appropriate actions are taken for the af Neted functional units.

c.

The Minimum Channels OPERABLE requirement is met; however, one additional channel may be removed from service for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, provided one of the inoperable channels is placed in the tripped condition.

MILLSTONE - UNIT 2 3/4 3-16 Amendment No. U p, 177, J M,

0303

l l

4 4

TABLE 3.3-3 (Continued)

ACTION 3 -

With less than the minimum channels OPERABLE the containment i

purge valves are to be maintained closed.

ACTION 4 -

With the number of '0PERABLE channels one less than the Total Number of Channels and with the pressurizer pressure:

a.

< 1750 psia: immediately place the inoperable channel in the bypassed condition; restore the inoperable channel to 4

OPERABLE status prior to increasing the pressurizer pressure above 1750 psia, b.

21750 psia, pperation may continue with the inoperable channel in the bypassed condition, provided the following condition is satisfied:

1.

The Minimum Channels OPERABLE requirement is met; however, one additional channel may be removed from service for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing l per Specification 4.3.2.1.1 provided BOTH of the inoperable channels are placed in the bypassed condition.

d

?

l l

4 NILLSTONE - UNIT 2 3/4 3-17 Amendment No. JJ7, J/7, Jpf, 0360

g5' TABLE 4.3-6

.p M.

REMOTE SHUTDOWN MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS g.

CHAf'NEL CHANNEL INSTRUMENT CHECK CALIBRATION g

4 1.

Wide Range Logarithmic M

R*

l Neutron Flux ro

'2.

Reactor Trip Breaker M

N.A.

Indication 3.

Reactor Cold Leg Temperature M

R 4.

Pressurizer Pressure a.

Low Range M

R b.

High Range M

R g

[

5.

Pressurizer Level M

R h

6.

Steam Generator Level M

R 7.

Steam Generator Pressure M

R

s(
  • Neutron detectors are excluded from the CHANNEL CALIBRATION.

5 F

- =

3/4.3 INSTRUMdWATIOJ BASES 3/4.3.1 AND 3/4.3.2 PROTECTIVE AND ENGINEERED SAFETY FEATUES (ESF) INSTRUMENTATION-The OPERABILITY of the protective and ESF instrumentation systems and bypasses ensure that 1) the associated ESF action and/or reactor trip will be initiated when the parameter monitored by each channel or combination thereof exceeds its setpoint, 2) the specified coincidence logic is maintained,

3) sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance, and 4) sufficient system functional capability is available for protective and ESF purposes from diverse parameters.

The OPERABILITY of these systems is required to provide the overall reliability, redundance and diversity assumed available in the facility design 1

for the protection and mitigation of accident and transient conditions. The integrated operation of each of these systems is consistent with the assumptions used in the accident analyses.

Action Statement 2 of Tables 3.3-1 and 3.3-3 requires an inoperable Reactor Protection System (RPS) or Engineered Safety Feature Actuation System (ESFAS) channel to be placed in the bypassed or tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The inoperable channel may remain in the bypassed condition for a maximum of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. While in the bypassed condition, the affected functional i

unit trip coincidence will be 2 out of 3.

After 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the channel must j

either be declared OPERABLE, or placed in the tripped condition.

If the channel is placed in the tripped condition, the affected functional unit trip coincidence will become 1 out of 3.

One additional channel may be removed from service for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, provided one of the inoperable channels is placed in the tripped condition.

i Plant operation with an inoperable pressurizer high pressure reactor protection channel in the tripped condition is restricted because of the j

potential inadvertent opening of both pressurizer power operated relief valves I

(PORVs) if a second pressurizer high pressure reactor protection channel i

failed while the first channel was in the tripped condition.

This plant operating restriction is contained in the Technical Requirements Manual.

The surveillance requirements specified for these systems ensure that the overall system functional capability is maintained comparable to the original design standards. The periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability.

The surveillance testing verifies OPERABILITY of the RPS by overlap testing of the four interconnected modules: measurement channels, bistable trip units, RPS logic, and reactor trip circuit breakers. When testing the measurement channels or bistable trip units that provide an automatic reactor trip function, the associated RPS channel will be removed from service, declared inoperable, and Action Statement 2 of Technical Specification 3.3.1.1 entered. When testing the RPS logic (matrix testing), the individual RPS channels will not be affected.

Each parameter within each RPS channel supplies three contacts to make up the 6 different logic ladders / matrices Y

(AB, AC, AD, BC, BD, and CD).

During matrix testing, only one logic matrix is tested at a time.

Since each RPS channel supplies 3 different logic MILLSTONE - UNIT 2 B 3/4 3-1 Amendment No. Jp/, Jpp, Jpp, 0305

~

m m

3/4.3 INSTRUNENTATION BASES 3/4.3.1 AND 3/4.3.2 PROTECTIVE AND ENGINEERED SAFETY FEATURES s

(ESF) INSTRUMENTATION fcontinued) i ladders, testing one ladder matrix at a time will not remove an RPS channel from the overall logic matrix. Therefore, matrix testing will not remove an l

RPS channel from service or cake the RPS channel inoperable.

It is not necessary to enter an action statement while performing matrix testing. This also applies when testing the reactor trip circuit breakers since this test will not remove an RPS channel from service or make the RPS channel inoperable.

l 1

The measurement of response time at the specified frequencies provides assurance that the protective and.ESF' action function associated with each channel is completed within the time limit assumed in the accident analyses.

No credit was taken in the analyses for those channels with response times indicated as not applicable. The Reactor Protective and Engineered Safety Feature response times are contained in the Millstone Unit No. 2 Technical Requirements Manual. Changes to the Technical Requirements Manual require a 10CFR50.59 review as well as a review by the Plant Operations Review Committee.

l The containment airborne radioactivity monitors (gaseous and particulate) i are provided to initiate closure of the containment purge valves upon detection of high radioactivity levels in the containment. Closure of these valves prevents excessive amounts of radioactivity from being released to the environs in the event t,f an accident. The actuation logic for this function is I out of 4.

Action Statement 3 of Table 3.3-3 addresses inoperable containment purge channels.

l l

l l

l l

l l

l

?

l l

MILLSTONE - UNIT 2 B 3/4 3-la Amendment No. 177, Jpp, Jpp, t

0305

l Docket No. 50-336 B17492 Millstone Nuclear Power Station, Unit No. 2 Proposed Revision to Technical Specifications Reactor Protective end Enginsered Safety Feature Actuation System Instrumentation History of the Millstone Unit No. 2 Technical Specifications for Page 2-4 from License Amendment No. 52 to License Amendment No.199 Page 2-5 from License Amendment No. 52 to License Amendment No. 61 November 1998 i

IABLE 2.2-1

~

REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS r-r-

M ALLOWABLE VALUES E

TRIP SETPOINT FUNCTIONAL UNIT Nct Applicable E

1.

Manual Reactor Trip Not applicable

~~

m 2.

Power Level - High e 9.88% above THERMAL POWER. with < 9.88% above THERMAL POWE Four Reactor Coolant Pumps a minimum setpoint of 5 15% of i minimurii setpoint. of s 15% of l

Operating RATED THERMAL POWER, and-a maximum RATED THERMAL POWER, and a maximum 107% of RATED THERMAL POWER.

of S 107% of RATED THERMAL POWER. of S 3.

Reactor Coolant Flow - Low (1)

> 91.7% of design reactor coolant

> 91.7% of design reactor coolant Four Reactor Coolant Pumps flow with 4 pumps operating

  • flow with 4 pumps operating
  • Operating

[

4.

Pressurizer Pressure - High 1 2400 psia 5 2400 p'sia 5.

Containment Pressure - High.

1 4.75 psig 5 4.75 psig

> 500 psfa 6.

Steam Generator Pressure -

-> 500 psia low (2)(5)

> 36.0% Water Level - each l

p 7.

Steam Generator Water Level-

-> 36.0% Water Level - each steam generator steam generator Low (5) g Trip setpoint adjusted to Local Power Density - High (3)

Trip setpoint adjusted to not exceed the lir?: lines not exceed the limit lines of Figures 2.2-1 and L.2-2.

g 8.

of Figures 2.2-1 and 2.2-2.

e

.E Thermal Margin / Low Pressure (1) g 9.

Four Reactor Coolant Pumps Trip setpoint adjusted to Trip setpoint adjusted to not exceed the limit ifne',

not exceed the limit lines Operating of Figures 2.2-3 and 2.2-4.

of Figures 2.2-3 and 2.E-4.

m 370,000 gpm

  • Design reactor :oolar.t flow with 4 pumps operating is q

,O, v...

L

-r m

TABLE 2.2-1 (Continued)

REACTOR PROTECTIVE INSTRL%1ENTATION TRIP SETPOINT LIM ALLOWABLE VALUES TRIP SETPOINT-r-

FUNCTIONAL UNIT-

> 500 psig El

> 500 psig

~

10. Loss of Turbine -- Hydraulic

. Fluid Pressure - Low (3)

> 829 rpm Z

11. Underspeed - Reacter

> 829 rpm

~

x

~

Coolant Pumps (1) m TABLE NOTATION Trip may be bypassed below 5% of RATED THERMAL POWER; bypass s

~

(1) when THERMAL POWER is > 5% of RATED THERMAL POWER.

Trip may be m'anually. bypassed below 600 ps:s; bypass'-shall be automatically rem (2) above 600 psfa.

[

Trip may be bypassed below 15% of RATO THERMAL POWER; bypass sh w

(3) when THERMAL POWER is >15% of RATED THERMAL POWER.

Each of four channels actuate on the auctioneered output of two transmitters, o (5) steam generator.

t i

er O

N

v

~

I TABLE 2.2-1 e

3 REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS

((

~

ALLOWABLE VALUES TRIP SETPOINT j

FUNCTIONAL UNIT 1.

Manual Reactor Trip Not Applicable-Not Applicable "z5 2.

Power Level-High.

< 9.7% above THERMAL POWER, with

< - 9.6% above THERMAL POWER,

~

Four Reactor Coolant Pumps s~f th a minfaum setpoint of a minimum of < 14.7% of. RATED Opera ting

< 14.6% of RATED THERMAL THERMAL POWER and a maximum of

~

< 106.7% of' RATED THERMAL POWER.

F0WER, and a maximum of

~

< 106.6% of RATED THERMAL V0 DER.

3.

Reactor Coolant Flow - Low (1)

> 90.1% of reactor coolant flow

> 91.7% of reactor coolant Four Reactor Coolant Pumps flow with 4 pumps operating *.

w'ith'4 pumps operating *.'

m i

Operating

> 823 rpm 4

Reactor Coolant Pump

> 830 rpm Speed - Low 5.

Pressurizer Pressure - High

< 2400 psia i 2408 psia 6.

Containment Pressure - High

< 4.75 psig

< 5.23 psig p

R

> 500 psia

~> 492 psia E

7.

Steam Generator Pressure -

~

Low (2) (5)

> 36.0% Water Level - each

> 35.2% Water Level - each steam 8.

Steam Generator Water Level - Low (5) steam generator genera tor z

Trip setpoint adjusted to not 9

Local Power Density - High (3)

Trip setpoint adjusted to not exceed the limit lines of exceed the limit lines of 2

Figures 2.2-1 and 2.2-2 (4).

Figures 2.2-1 and 2.2 2 (4).

l ta

b Design Reactor Coolant flow with 4 pumps operating is 370,000 gpm..

a a

2 m-

v TABLE 2.2-1 iContinued),

m x

REACTOR PROTECTIVE INSTRUMENTATI'ON TRIP SETPOINT LIMITS..

8 ALLOWABLE VALUES TRIP SETPOINT M FUNCTIONAL UNIT E

10.

Thermal !!argin/ Low Pressure (1)

Trip setpoint adjusted to not.

[

Four Reactor Coolant Pumps Trip setpoint adjusted to not exceed the limit lines of exceed the limit lines of

/'

Operating Figures 2.2-3 anij 2.2-4 (4).

Figures 2.2-3 and 2.2-4 (4).

11.

Loss of Turbine -- Hydraulic 3 500 psig 1 500 psig Fluid (3) Pressure - Low TABLE NOTATION Trip 'may be bypassed below 5% of RATED THERf!AL POWER: bypass shall be automatically 7

(1) removed when TilERt1AL POWER is,; 5% of RATED tiler'fAL POWER.

Trip may be manually bypassed below 600 psia; bypass shall be automatically removed (2) at or above 600 psia.

Trip may be bypassed below 15% of RATED THERMAL POWER; bypass si all be automatically h

j (3) 15% of RATED THERMAL POWER.

removed when THERMAL POWER is 3

-~

Calculations of the trip setpoint includes measurement, calculational and processor g

(4) uncertainties, and dynamic allowances,

=

Each of four channels actuate on the auctioneered output of two transmitters, one from 3

(5) each steam generator.

.E E

e

t TABLE 2.2-1

.."E.

}

REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LI!!!TS E

FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES m

l.

Manual Reactor Trip Not Applicable

.Not Applicable-E7 2.

Power Level-High Four Reactor Coolant Pumps

< 9.6% above THERMAL-POWER,

< 9.7% above THERMAL POWER, with m-Operating iiith a minimum setpoint of a minimum of < 14.7% of. RATED g

< 14.6% of RATED THERMAL THEllMAL POWER, and a nsaximum of F0WER, and a maximum of

< 106.7% of RATED THERMAL POWER.

< 106.6% of RATED THERMAL POWER.

t 3.

Reactor Coolant Flow - Low (1)

Four Reactor Coolant Pumps

> 91.7% of reactor coolant

> 90.1% of reactor coolant flow i

Operating T1ow with 4 pump operating *.

~ith 4 pumps operating *.

m 4.

Reactor Coolant Pump

> 830 rpm

> 823 rpm Speed - Low 5.

Pressurizer Pressure - Hiah

< 2400 psia

< 2408 psia 6.

Containment Pressure - High

< 4.75 fo

< 5.23 psig p

7.

Steam Generator Pressure -

> 500 psi /

> 492 psia

^"

~

k Low (2) (5)

~

'A t

8.

Steam Generator Water

> 36.0% Water Level - each

> 35.2% Water Level - each steam 2

?

Level - Low'(5) steam generator generator b

9.

Local Power Density - High (3)

Trip setpoint adjusted to not Trip setpoint adjusted to not exceed the limit lines of exceed the ifmit lines of Figures 2.2-1 and 2.2-2 (4).

Figures 2.2-1 and 2.2-2 (4).

s.

{

'T Design Reactor Coolant flow with 4 pumps operating is 362,600 gpm.

I

.. ~..

~h

,e s

. TABLE 2e2-1 1

REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LI!iITS_

E~

N..

TRIP'SETPOINT ALLOWABLE VALUES E

FUNCTIONAL-UNIT 1.

Manual Reactor Trip Not Applicable Not Applicable Q

2.

Power Level-High Four Reactor Coolant Pumps

< 9.6% above THERNAL POWER,.

<-9.7% above THERMAL POWER, with' i

w with a minimum setpoint of a minimum of 1 14.7% of RATED Operating

< 14.6% of RATED THEP. MAL THERMAL POWER, and a maximum of POWER, and a makimum of

< 106.7%.of RATED THERMAL POWER.

< 106.6% of RATED THERMAL POWER.

3.

Reactor Coolant Flow - Low (1)

Four Reactor Coolant Pumps

> 01.7% of reactor coolant

> 90.1% of reactor coolant' flow.

flow with 4 pumps operating *.

with 4 pumps operating *.

m 1

Operating 4.

Reactor Coolant Pump

> 830 rpm

> 823 rpm Speed - Low 5.

Pressurizer Pressure - High

< 2400 psia 1 2408 psia 6.

Containment Pressure - High

< 4.75 psig

< 5.23 psig R

R 7.

Steam Generator Pressure -

> 500 psia

> 492 psia

~

i

~

~

Low (2) (5)

?

S.

Steam Generator Water

> 36.0% '!ater Level - each

> 35.2% Water Level - each steam Level - Low (5) steam generator generator 2

L h

9.

Local Power Density - High (3)

Trip setpoint adjusted to not

~ Trip setpoint adjusted to not exceed the limit lines of exceed the limit lines of Figures 2.2-1 and 2.2-2 (4).

Figures 2.2-1 and 2.2-2 (4).

+

so b

I

&~

I Design Reactor Coolant f. low with 4 pumps onerating is 350,000 gpm.

k o

r

-a

=

.e.

=

m.

<.c..-

=

..-e m

- -.,_ __., -.:., m. :x:e tmm.

m...__

..._a,,~--.~

- -,.- m m er( ' m u m m am n

2

,,,,,,,,,cm..,

' T.hk#

p TABLE 2.2-1

.2 P

REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LI!!ITS' S

E FUNCTIONAL UNIT TRIP 5ETPOINT

. ALLOWABLE. VALUES

[

1.

Manual Reactor Trip.

~

Not Applicable Not Applicable z

1 2.

Power Level-High N

Operating

~ith a minimum setp'oint of

~< 9.7% above THERMAL POWER. with Four Reactor Coolant Pumps

< 9.6% above THERMAL POWER, minimum of < 14.7% of RATED

< 14.6% of RATED THERMAL THERMAL POWER 7 and a maximum of F0WER, and a maicimum of

< 106.7% of RATED THERMAL POWER.

~

< 106.6% of RATED. THERMAL POWER.

f 3.

Reactor Coolant Flow - Low (1)

Four Reactor Coolant Pumps

> 91.7% of reactor coolant

> 90.1% of reactor coolant flow

~

i Operating fie wi+b 4 pumps operating *.

~ith 4 pumps operating *.

4.

Reactor Coolant Pump

,_ 830 rpm

> 823 rpm Speed - Low 5.

Pressurizer Pressure - High

< 2400 psia

< 2408 psia E

6.

Containment Pressure - High

< 4.75 psig

< 5.23 psig 5

l A

7.

Steam Generator Pressure -

~> 500 psia

~> 492 psia Low (2) (5)

=

?

Level - Low (5)

~> 36.0% Water Level - each

~> 35.2% Water Level - each steam 8.

Steam Generator Water

~

team generator enera tor g

i w

9.

Local Power Density - High (3)

Trip setpolnt a,djusted to not Trip setpoint adjusted to not

?

. exceed the limit, lines of exceed the limit lines of Figures 2.2-1 and 2.2-2 (4).

Figures 2.2-1 and 2.2-2 (4).

4

?

I e

Design Reactor Coolant flow with 4 pumps operating is' 340,C30 gpm.

w e

h d

Tsu E 2.2-1 REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LTHlTS x

FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES 1.

Manual Reactor Trip Not Applicable,

Not Applicable.

2.

Power Level-High

.z U

Four Reactor Coolant Pumps s 9.6% above THERMAL POWER, s 9.7% Above THERMAL POWER,.

Operating with a minimum setpoint of with a minimum of 514.77.

~

s 14.6% of RATED THERMAL of RATED THERMAL POWER, and

~

w a maximum of < 106.7% of

% POWER.

f

's RATED THERMAL POWER.

3.

Reactor Coolant Flow -

low (1) 2 91,7% cf reactor coolant l2 90.1% of reactor coolant flow with 4 pumps operating *.

with 4 pumps operating.

4.

Reactor Coolant Pump 2 830 rpm 1 823 rpm Speed - Low a

5.

Pressurizer Pressure - High 5 2400 psia 5 2408 psia Q [

6.

Containment Pressure - High 5 4.75 psig f 5.24 psig s

[

7.

Steam Generator Pressure -

2 680 psia 2'672 psia g

Low (2)

(5)

F 8.

Steam Generator Water 2 36.0% Water Level - each 2 35.2% Water Level - each

(

Level - Low (5) steam generator steam generator m(

9.

Local Power Density -

Trip setpoint adjusted to not Trip setpoint adjusted to High (3) exceed the limit lines of not exceed the limit lines a(

Figures 2.2-1 and 2.2-2 (4).

of Figures 2.2-1 and y

., 2.2-2 (4).

2

  • Design Reactor Coolant flow with 4 pumps operating is the lesser of either:

a.

The reactor coolant flow rate measured per specification 4.2.6.1, or

,y b.

340,000 gpm U

u m

TAB W

v REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS v

3 FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES 1.

Manual Reactor Trip Not Applicable Not Applicable z

'{

2.

Power Level-High Four Reactor Coolant Pumps S 9.6% above THERMAL POWER, S 9.7% Above THERMAL POWER, with a minimum setpoint of with a minimum of S 14.7%

of' RATED THERMAL POWER, and Z

Operating

$ 14.6% of RATED THERMAL a maximum of f 106.7% of POWER.

~

RATED THERMAL POWER.

3.

Reactor Coolant Flow -

2 91.7% of reactor coolant 2 90.1% of reactor coolant Low (1) flow with 4. pumps operating *.

with 4 pumps operating.

4.

Reactor Coolant Pump 2 830 rpm 2 823 rpm Speed - Low

[

5.

Pressurizer Pressure - High S 2400 psia 5 2408 psia 6.

Containment Pressure - High S 4.75 psig 5 5.24 psig h

7.

Steam Generator Pressure -

2 680 psia 2 672 psia a

low (2)

(5)

N 8.

Steam Generator Water 2 36.0% Water Level - each 2 35.2% Water Level - each NB Level - Low (5) steam generator steam generator

. _. 2 9.

Local Power Density -

Trip setpoint adjusted to not Trip setpoint adjusted to kO exceed the limit if nes of not exceed the limit lines 3a High (3)

Figures 2.2-1 and 2.2-2 (4).

of Figures 2.2-1 and 2.2-2 (4).

P

  • Design Reactor Coolant flow with 4 pumps operating is the lesser of either:

The reactor coolant flow rate measured per specification 4.2.6.1, or i

a.

The minimum value specified in the CORE OPERATING LIMITS REPORT.

y y

b.

TABLE 2.2-1 REACTOR PROTECTIVE _ INSTRUMENTATION TRIP SETPOINT LIMITS o,5f 11 ALLOWABLE VALUES TRIP SETPOINT k

FUNCTIONAL UNIT Not Appifcable 1.

Manual Reactor Trip Not Appitcable E

Power Level-High f 9.7% Above THERMAL POWER, 2.

q 3 9.6% above THER?t4L POWER, with a minimum of 514.7%

Four Reactor Coolant Pumps with a minimum setpoint of of RATED THERMAL POWER, and a N

Operating s.14.6% of RATED THERMAL maximum.of f 106.7% of POWER.

RATED THERMAL POWER.

2 91.7% of reactor coolant 2 90,9% of reactor coolant flow 3.

Reactor Coolant Flow -

flow with 4 pumps operating *.

with 4 pumps operating.

Low (1) 2 823 rpm 4.

Reactor Coolant Pump 2 830 rpm Speed - Low

$ 2408 psia Pressurizer Pressure - High S 2400 psia l

S.

S 5.24 psig Containment Pressure - High

$ 4.75 psig l

6.

2 672 psia 7.

Steam Generator Pressure -

2 680 psia y

Low (2)

(5) y, 2 36.0% Water Level - each 2 35.2% Water Level - each steam generator Ok 8.

Steam Generator Water steam generator Level - Low (5)

?g Trip setpoint adjusted to not Trip setpoint adjusted to not exceed the limit lines M ac 9.

Local Power Density -

exceed the limit lines of

?*

High (3)

Figures 2.2-1 and 2.2-2 (4).

of Figures 2.2-1 and 2.2-2 (4).

p l

(%

z N

?

g r

  • Design Reactor Coolant flow with 4 pumps operating is the lesser of either:

e

a. The reactor coolant flow rate measured per Specification 4.2.6.1, or 5
b. The minimum value specified in the CORE OPERATING LIMITS REPORT.

w?

M

.