Text

Amend 60 to License NPF-49,revising Tech Specs to Support Refueling & Cycle 4 Operation of Facility Using Vantage 5 Hybrid Fuel Design
	ML20070K460
Person / Time
Site:	Millstone
Issue date:	03/11/1991
From:	Stolz J; Office of Nuclear Reactor Regulation
To:	; Northeast Nuclear Energy Co (NNECO)
Shared Package
ML20070K463	List: ML20070K460; ML20070K490; ML20070K512;
References
	NPF-49-A-060 NUDOCS 9103180439
	Download: ML20070K460 (117)
	v • d • e

8(gies asog jo, UNITED 5i ATES g

NUCLEAR REGULATORY COMMISSION e

W ASHING TON, D. C. 20666

...+#

tt0RTHEAST NUCLEAR ENERGY _C0fiPANY,_ET AL.

DOCLET NO. 50-423 MILLSTONE NUCLEAR POVER STATION, UNIT NO. 3 AMEN 0 MENT TO FACILITY GPERATING LICENSE Amendment No. 60 License No. NPF-49 1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The app (lications for amendment by Northeast Nuclear Energy Company, et al.

thelicensee)datedOctober 25,1990(supplementedFebruary 11,1991), and November 1, 1990 (supplemented November 2, 1990, November 30, 1990, December 4,1990, February 15, 1991 and February 22, 1991) comply with the standards and requireroents of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

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

There is reasonable 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 Comission's regulations; i

D.

The issuance of this amendment will not be inimical to the comon l

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 regulations and all applicable requirements have i

been satisfied.

l l

9103100439 910311 DR ADOCK 0500 3

I l

2-i 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 Operatin9 License No. HPF-49 is hereby amended to read as follows:

i (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 60 and the Environmental Protection Plan contained in Appendix B both of-which are attached hereto are l

hereby incorporated in the license. The licensee shall operate the 1

facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This. license amendment is effective as of the-date of'its issuance, to be implemented within 30 days of issuance.

FOR THE NUCLEAR REG LATORY COMMISSION C

Jo/jectDirectorate1-4 n F. Stoir, Dire r

Po ivision of Reactor Projects - 1/11-l Office of Huclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

March 11, 1991 i

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4r ATTACHMENT TO LICENSE AMENDMENT NO. 60 FACILITY OPERATING LICENSE NO. NPF-49 DOCKET NO.00-423 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages, The revised pages are identified by amendment number and contain vertical lines indicating the areas of change, Remove Insert ii ii iv iv v

v xiii xiii 1-5 1-5 1-7 1-7 2

2-2 2-2 2-3 2-3 2-5 to 2-11 2 2-12 B 2-1 0 2-1 B25 B 2-5 B 2-7 B 2-7 3/41-1 to 1-23 3/4 1-1 to 1-27 3/4 2-1 to 2-24 3/4 2-1 to 2-28 3/4 3-2 3/4 3-2 3/4 3-4 to 3-6 3/4 3-4 to 3-6 3/4 3-8 3/4 3-8 3/4 3-12 to 3-14 3/4 3-12 to 3-14 3/4 3-26 3/4 3-26 3/4 4-2 3/4 4-2 3/4 4-6 3/4 4-6 to 4-6a 3/4 4-8 3/4 4-8 3/4 4-30 3/4 4 No change 3/4 4-34 3/4 4 No change 3/4 4-35 3/4 4 No change 3/4 4-40 3/4 4 No change 3/4 4-41 3/4 4 No change 3/4 5-1 3/4 5-1 3/4 5-4 to 5-6 3/4 5-4 to 5-6 3/4 5-9 3/4 5-9 3/4 6-14 3/4 6-14 3/4 9-1 3/4 9-1 B 3/4 1-1 B 3/4 1-1 i

B 3/4 1-3 to 1-4 B 3/4 1-3 to 1-4 B 3/4 2-1 to 2-7 8 3/4 2-1 to 2-6 B 3/4 4-1 B 3/4 4-1 to 4-14 B 3/4 9-1 B 3/4 9-1 5-6 5-6 6-21 6-21 6-21a 6-21a

. ~. ~

h INDEX DEFINITIONS

.CTION P.AGI

.32 SLAVE RELAY TEST.............................................

1-6 1.33 SOURCE CHECK.................................................

1-6 1.34 STAGGERED TEST BASIS.........................................

1-6 1.35 THERMAL P0WER................................................

16 1.3d TRIP ACTUATING DEVICE OPERATIONAL TEST.......................

1-6 1.37 UNIDENTIFIED LEAKAGE.........................................

1-6 1.34 UNRESTRICTED AREA............................................

1-6 1.39 VENTING......................................................

1-7 1.40 SPENT FUEL POOL STORAGE PATTERNS.............................

17 1.41 SPENT FUEL POOL STORAGE PATTERNS.............................

1-7 1.42 CORE OPERATING LIMITS REP 0RT.................................

1-7 1,43 ALLOWED ?0WER LEVEL-APLND,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,

3,7 1,44 ALLOWED POWER LEVEL--APL0l...................................

1-7 TABLE 1.1 FREQUENCY NOTAT!0N......................................

18 TABLE 1.2 OPERATIONAL M0 DES.......................................

1-9 l

l MILLSTONE - UNIT 3 ii Amendment No JS, EE, 60 l

0012

A l@ll i

LIMITING CONDITIONS FOR OPERATION AND SVRVElllANCE RE0VIREMENTS i

i SECTION MG1 3/4.0 APPLICABILIJJ..............................................

3/4 0 1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORAT10N CONTROL Shutdown Margin - MODES 1 AND 2.........................

3/4 1-1 Shutdown Margin MODES 3, 4. AND 5 LOOPS FILLED........

3/4 1 3 FIGURE 3.1 1 REQVIRED SHUTDOWN MARGIN FOR MODE 3 WITH FOUR LOOPS IN OPERATION...............

3/4 1 4 FIGURE 3.1 2 REQUIRED 3HUTDOWN MARGIN FOR MODE 3 WITH THREE LOODS IN OPERATION..............

3/4 1 5

. 1 FIGURE 3.1 3 REQUIRED SHUTDOWN MARGIN FOR MODE 4........

3/4 1-6 FIGURE 3.1-4 REW'JIRED SHUTDOWN MARGIN FOR MODE 5 WITH RCS LOOPS FILLED..........................

3/4 1-7 Shutdown Margin - Cold Shutdown -

Loops Not filled...........................

3/4 1 8 FIGURE 3.1 5 REQUIRED SHUTDOWN MARGIN FOR MODE 5 WITH RCS LOOPS DRAINED...........................

3/4 1 9 Moderator Temperature Coefficient.......................

3/4 1-10 Minimum Temperature for Criticality......................

3/4 1-12 3/4.1.2 BORAT10N SYSTEMS Flow Path - Shutdown....................................

3/4 1-13 Flow Paths - Operating..................................

3/4 1 14 Charging Pump Shutdown................................

3/4 1 15-Charging Pumps - Operating..............................

3/4 1-16 Borated Water Source - Shutdown.........................

3/4 1-17 Borated Water Sources - Operating.......................

3/4 1-18 3/4.1.3 MOVABLE CONTROL ASSEMBLIES-Group Height............................................

3/4 1-20 TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN INOPERABLE FULL LENGTH R00..................

3/4 1-22

-Position Indication Systems - Operating.................

3/4 1 23-l I

MILLSTONE - UNIT 3 iv Amendment No JJ 60 0039 L

-~

--~

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

_. _ _ ~ _ _ _.

INDEX LIMITING CONDITIONS FOR OPERATION AND SVRVEILLANCE RE0VIREMERIS i

SECTION 12E Position Indication System Shutdown....................

3/4 1 24 Rod Drop Time............................................

3/4 1-25 Shutdown Rod Insertion Limit.............................

3/4 1-26 Control Rod insertion Limus.............................

3/4 1 27 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX 0!FFERENCE....................................

3/4 2 1 Four Loops 0perating.....................................

3/4 2 1 Three Loops 0perating....................................

3/4/2 4 i

3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR F

(Z).....................

3/4 2 7 g

Four Loops 0perating.....................................

3/4 2-7 Three loops Operating.....................................

3/4 2 11 1

3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACT0R...................................................

3/4-2-15 Four Loops 0perating.....................................

3/4 2-15 Three loops 0perating....................................

3/4 2-18 3/4.2.4 QUADRANT POWER Tl'.T RAT10................................

3/4 2 20 3/4.2.5 DNB PARAMETERS...........................................

3/4 2 23 TABLE 3.2 1 DNB PARAMETERS........................................

3/4 2 24 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION......................

3/4 3 1 TABLF 3.3 1 REACTOR TRIP SYSTEM INSTRUMENTATION....................

3/4 3-2 TABLE 3.3 2 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES....3/4 38 TABLE 4.3 1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVE REQUIREMENTS................................ILLANCE' 3/4 3-10 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM i

INSTRUMENTATION..........................................

3/4 3-15 TABLE 3.3 3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION..........................................

3/4 3 17 TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETP0lNTS...........................

-26 i

M!l.LSTONE - UNIT 3 v

Amendment.No. 59_60

I INVEX BASES SECTION P.Mf, 3/4.0 APPLICABILITY...............................................

B 3/4 0 1 3/4.1 REACTIVITY CONfROL SYSJrg 3/4.1.1 B3 RATION CONTROL................

B 3/4 1.......................

3/4.1.2 "8dTION SYSTEMS...........

B 3/4 1-2 3/4.1.3

. CONTROL ASSEMBLIES....,...........................

B 3/4 1 3 3/4.2 "y2 'l

?!BUTIONLIMITi...................................

B 3/4 2-1 3/4.2.1 W! \\L

.UX DIFFERENCE.....................................

B 3/4 2-1 3/4.2.2 ant

/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and RCS FLOW-RATS \\ND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR,........

B 3/4 2-3 3/4.2.4 QUADRANT POWER TILT _RATI0.................................

B 3/4 2-5 3/4.2.5 DNB PARAMETERS............................................ 'B 3/4 2-5 3/4.3 INSTRUMENTATION 3/4.3.1 and-3/4.3.2 REACTOR TRIP SYSTEM INSTRUMENTATION and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION,.........................n............_.,

B 3/4 3-1

-3/4.3.3_ MONITORING INSTRUMENTATION................................

B 3/4 3-3 p

3/4.3.4 TURBINE OVERSPEED PROTECTION...............................

B_3/4 3-7 3/4.4 REACTOR C00UNT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND-COOLANT. CIRCULATION............ B 3/4 4-1 3/4,4.2 SAFETY VALVES.............................................

B 3/4 4-2 3/4.4.3 PRESSURIZER...............................................

B 3/4 4-2 3/4.4.4 RELIEF VALVES.............

B 3/4 4-2 3/4.4.5 STEAM GENERATORS......................................_.... 9 3/4 4-3 3/4.4.6 REACTOR-COOLANT SYSTEM LEAKAGE............................

B 3/4-4-4 3/4.4.7 CHEMISTRY.................................................

B 3/4 4-5 3/4.4.B SPECIFIC ACTIVITY.........................................

B 3/4 4-5 3/4.4.9 PREESURE/ TEMPERATURE LIMITS...............................

B 3/4 4-7 l

MILLSTONE -' UNIT 3 xiii Amendment No. 60 l

'0014

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

r q

DEFINITIONS OVADRANT POWER TILT RATIO 1.24 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector cali-brated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.

With one excore detect (* inoperable, the remaining three detectors shall be used for computing the aurage.

RADIOACTIVE WASTE TREATMENT SYSTEMS 1.25 RADI0 ACTIVE WASTE TREATMENT SYSTEMS are those liquid, gaseous and solid waste systems which are required to maintain control over radioactive material in order to meet the Limiting Conditions for Operation (LCOs) set forth in these specifications.

RADIOLOGICAL EFFLUENT MONITORING AND OFFSITE DOSE CALCULATION MANUAL (REMODCM) 1.26 A RADIOLOGICAL EFFLUENT MONITORING MANUAL shall be a manual containing the site and environmental sampling and analysis programs for measurements of radiation and radioactive materials in - those exposure pathways and for those radionuclides which lead to the highest potential radiation exposures to in-dividuals from station operation.

An OFFSITE DOSE CALCULATION MANUAL shall be a manual containing the methodology and parameters to be used in the calcula-tion of offsite doses due to radioactive gaseous and liquid effluents and in the calculation of gaseous and liquid afluent monitoring instrumentation alarm / trip setponts.

Requirements of t'r IMODCM are provided in Specifica-tion 6.13.

RATED THERMAL POWER 1.27 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3411 MWt.

REACTOR TRIP SYSTEM RESPONSE TIME 1.28 The REACTOR TRIP -SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its Trip Setpoint at the channel sensor until loss of stationary gripper coil voltage.

?

REPORTABLE EVENT l

l.29 A REPORTABLE EVENT shall be-any of those conditions specified in Section 50.73 of 10 CFR Part 50.

SHUTOOWN MARGIN

~

1.30 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which t

the reactor is subcritical or would be subcritical-from its present condition assuming all full-length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest j

reactivity worth which is as m ed to be fully withdrawn.

gSTONE-UNIT 3 1-5 Amendmont No. 60 c

-.c

-c w

-.e e

.=w-.are

r+~e

-e

.ss-

=

+

y:

DEFINITIONS VENTING 1,39 VENTING shall be the controlled process of discharging air or gas from a cor.finement to maintain temperature, pressure, humidity, concentration, or other operating condition, in such a manner that replacement air or gas is not provided or required during VENTING.

Vent, used in system names, does not imply a VENTING process.

SPENT FUEL POOL STORAGE PATTERNS:

1.40 Region I spent fuel racks contain a cell blocking device in every 4th location for criticality control.

This 4th location will be referred to as the blocked location.

A STORAGE PATTERN refers to the blocked location and all adjacent and diagonal Region I cell locations surrounding the blocked location.

Boundary configuration between Region I and Region 11 must have cell blockers positioned in the outermost row of the Region I perimeter, as shown in Figure 3.9 2.

1,41 Region 11 contains no cell blockers.

CORE OPERATING LlHITS REPORT (COLR) 1.<2 The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document that provides core operating limits for the current operating reload cycle.

These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.6.

Unit Operation within these operating limits is addressed in individual specifications.

ALLOWED POWER LEVEL HO 1,43 APL is the minimum allowable nuclear design power level for base load operation and is specified in the COLR.

OL 1.44 APL is the maximum allowable power level when transitioning into base load operation.

MILLSTONE - UNIT 3 1-7 Amendment No. JJ, Jp, 60 0002

n, e

680

~

UNACCEPTABLE 1

660

^

2425 PSIA

-?

2250 PSIA i

a 640

~

d

~

i i

E 5

I 2000 PSIA i

~

N 620 i

1860 PSIA i

C I

c L

600 i

ACCEPTABLE P

OPERATION 580

~

=

i:

560

~

0 0.2 0.4 0.6 0.8 1.0 1.2 FRACTION OF RATED THERMAL POWER i

FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION WILLSTONE - UNIT 3 2-2 Amendment No. 60

I i

680 UNACCEPTABLE i

660 h OPERATION 2425 PSIA q

2250 PSIA i

^

i a

640

-i d

i E

5

~

2000 PSIA 620

~

m\\

5 i

m 1860 PSIA i:

E y

E

.4 600

-i l

ACCEPTARLE E

OPERATION

(

580 E

t 560 l

0 0.2 0.4 0.6 0.8 1.0 l

FRACTION OF RATED -THERMAL POWER.

FIGURE 2.1-2 REACTOR CORE SAFETY LIMIT - TWEE LOOPS IN OPERATION-WILLST0E - UNIT 3 2-3 Amendment No. 60

I TABLE 2.2-1 om REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS P

~ 0; TOTAL SENSOR g

ALLOWANCE ERROR 2

FUNCTIONAL UNIT (TA)

Z (S)

TRIP SETPOINT ALL(SABLE VALUE 1

l.

Manual Reactor Trip N.A.

N.A.

E Q

2.. Power Range, Neutron Flux

[

w a.

High Setpoint

1) Four Loops Operating 7.5 4.56 0

$ 109% of RTP**

$ 111.1% of RTP**

1

2) 1hree Loops Operating 7.5 4.56 0

s 80% of RTP**

$ 82.1% of RTP**

b.

Low Setpoint 8.3 4.56 0

s 25% of RIP **

127.1% of RTP**

3.

Power Range, Neutron Flux, 1.6 0.5 0

5 5% of RTP** with 16.3% of RTP** with High Positive Rate a time constant a time constant t

7, 2 2 seconds 2 2 seconds j

u 4.

Power Range, Neutron Flux, 1.6 0.5 0

s 5% of RTP** with 5 6.3% of RTP** with High Negative Rate a time constant a time constant

[

1 2 seconds 2 2 seconds i

5.

Intermediate Range, 17.0 8.41 0

1 25% of RTP**

-< 30.9% of RTP**

i Neutron Flux 1

)

{

6.

Source Range, Neutron Flux 17.0

.10.01 0

$ 10+5 cps

< l.4 x 10+5 cp3 o

7.

Overtemperature AT a.

Four Loops Operating x

1) Channels I, II 10.0 6.80 1.71 + 1.33 See Note 1

.See Note 2 g

(Temp + Press) l 8

2) Channels III, IV 10.0 5.83 1.71 + 2.60 See Note 1 See Note 2 (Temp + Press)

- RTP - RATED THERMAL POWER

)

l 1

2 \\

TABLE 2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS P

- r-U TOTAL SENSOR E

ALLOWANCE ERROR 7

FUNCTIONAL UNIT (TA)

I (S)

TRIP SETPOINT ALLOWABLE VALUE b.

Three Loops Operating

1) Channels I, II 10.0 6.80 1.71 + 1.33 See Note 1 See Note 2 w

(Temp + Press)

2) Channels III, IV 10.0 5.83 1.71 + 2.60 See Note 1 See Note 2 (Temp + Press) 8.

Overpower AT 4.8 1.24

'I.71 See Note 3 See Note 4 9.

Pressurizer Pressure-Low 5.0 1.77-3.3 2 1900 psia 1 1890 psia 10.

Pressurizer Pressure-High 5.0 1.77' 3.3 5 2385 psia 5 2395 psia 11.

Pressurizer Water Level High 8.0 5.13 2.7 5 89% of instrument 5.90.7% of instrument-span span 12.

Reactor Coolant Flow-Low 2.5 1.52 0.78 190% of loop 2 89.1% of loop design flow *

' design flow

E e

13. Steam Generator Water 18.10 16.64 1.50 218.10% of narrow

> 17.11% of narrow E.

Level Low-Low range instrument range instrument

-l span span

14. General Warning Alarm N.A.

N.A.

g 15.

Low Shaft Speed - Reactor 3.8 0.5 0

2 95.8% of rated 2 92.5% of rahd

' Coolant Pumps speed speed e

g

Minimum Measured Flow !%- Loop - 96,870 gpa (Four Loops Operating); 101,066 gpm (Three Loops Operating)

~

-3.

1

TABLE 2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUNENTATION TRIP SETPOINTS 1

2 Pl TOTAL SENSOR l

ALLOWANCE ERROR i

FUNCTIONAL UNIT (TA)

Z (S)

TRIP SETPOINT ALLOWABLE VALUE a

r E

16. Turbine Trip

~

.Z w

a.

Low Fluid Oil Pressure N.A.

N.A.

2 500 psig 2 450 psig b.

Turbine Stop Valve N.A.

N.A.

2 1% open 2 1% open Closure i

17. Safety Injection Input N.A.

N.A.

H.A.

from ESF r

18.

Reactor Trip System Interlocks

.10

_11 a.

Intermediate Range

.N.A.

N.A.

21 x 10 amp 2 6 x 10 amp 7

Neutron Flux, P-6

~

b.

Low Power Reactor Trips Block, P-7 1)

P-10 input.

N.A.

H.A.

N.A.

I 10% of RTP**

< 12.1% of RTP**

2)

P-13 input N.A.

N.A.

5 10% RTP** Turbine s 12.1% RTP** Turbine Impulse Pressure Impulse Pressure p

Equivalent Equivalent h

c.

Power Range Neutron j

Flux, P-8 a

g 1)

Four Loops Operating N.A.

N.A.

5 37.5% of RTP**

$ 39.6% of RTP**

~

2) Three Loops Operating N.A.

N.A.

5 37.5% of RTP**

1 39.6% of RTP**

- RTP = RATED THERMAL POWER 4

i

=

r TABLE 2.2-1 (Continued) l REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS or 2;

4

- c-c)

TOTAL SENSOR E

ALLOWANCE ERROR i

FUNCTIONAL UNIT (TA)

Z (S)

TRIP SETPOINT ALLOWABLE VALUE c

5 d.

Power Range Neutron N.A.

-N.A.

N.A.

5 51% of RTP**

1 53.1% of RTP**

Flux, P-9 w

e.

Power' Range Neutron N.A.

N.A.

N.A

t 10% of RTP**

> 7.9% of P.TP**

l Flux, P-10 i

t

19. Reactor Trip Breakers.

N.A.

N.A N.A.

.N.A.

{

l l

20. Automatic Trip and Interlock N.A.

N.A.

Logic 21.

Three Loop Operation N.A.

N.A.

-N.A.

N.A I

i Bypass Circuitry ro 4

i

+

i i

t t

.I l

3 n.

i 3

a 4-D t

I f

- RTP = RATED THERMAL POWER t

I I

-l l

5 i

t o3 TABLE 2.2-1 (Continued)

~Gg TABLE NOTATIONS 2

l 7

NOTE 1: OVERTEMPERATURE AT AT (1 + 7;S) 1 IKI~KI 54 ) [T I

+

S) <AT 1 + 7 3) - T'] + K3 (P - P') - f (AI)}

O 2 (1 + 7 5) g 3

]

(1 + 7 S)

I+#

2 3

5 6

Where:

AT

- Measured AT by Reactor Coolant System Instrumentation; I+#3 1

- Lead-lag compensator on measured AT; I+7S2 73, 72

- Time constants utilized in lead-lag compensator for AT, 73-8s, 2 - 3 s; j

7 I

- Lag compensator on measured AT; 1+r33

'?

3

- Time constants utilized in the lag compensator for AT, 73 - O s; j

e 7

AT

= Indicated AT at RATED THERMAL POWER; O

j

}

K

= 1.20 (Four_ Loops Operating); 1.20 (Three Loops Operating);

3 X

- 0.02456,

[

2 I+#S' l

[

= The function generated by the lead-lag compensator for T,yg dynamic 4

I'~+ # 5 h-5 compensation;

- Time constants utilized in the lead-lag compensator for T,yg, 74 - 20 s, 74,,75 5'" 4'83

- g 7

g

.T Average temperature, *F;

=

8 I

- Lag compensator on measured T,yg; i

1+r36

- Time constant utilized in the measured T,yg lag compensator, 7

7 6 - O s; 6

1.

d ox TABLE 2.2-1 (Continued)

~Gg TABLE NOTATIONS (Continued) z 7

NOTE 1:

(Continued)

T'

$ 587.l*F (Nominal T,yg at RATED THERMAL POWER);

6' K

0.001311/ psi; 3

P Pressurizer pressure, psia;

=

P' 2250 psia (Nominal RCS operating pressure);

S Laplace transform operator, s 2; i

and f powerl(AI) is a function of the indicated difference between top and bottom detectors of the range neutron ion chambers; with gains to be selected based on measured instrument response rp during plant startup tests such that:

(1) for q

- gb between -26% and + 3%, f (AI) = 0, where q and qb+are percent RATED THERMAL POWE in thb top and bottom halves of the core respectively,t 3

and g A

4 is total THERMAL. POWER in t

b percent of RATED THERMAL POWER; (2)

For each percent that the magnitude of q

-q exceeds -26%, the AT Trip Setpoint shall be automatically reduced by 3.55% of its value aI RATED THERMAL POWER; and F

(3)

For each percent that the magnitude of q - q exceeds +3%, the AT Trip Setpoint shall be 3

automatically reduced by 1.98% of its value ak RATED THERMAL POWER.

~

4 8

NOTE 2:

The channel'.s taximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 2.7%

y AT span (Four Loop Operation); 2.7% AT span (Three Loop Operation).

8 i

N

t L

02 a;--

TABLE 2.2-1 (Continued)

~Gg-TABLE NOTATIONS (Continued) l

[

NOTE 3: OVERPOWER A8T z

AT (1 + 7 S) (.

1.

.) 1 AT. (K. - K I#3 I

I I

) T - K. [T I I

) - T"] - f (AI))

[

~

3 7

s 2

(1 + 7 S) (1 + T S)

(1 + 7 5)

(1 + T S)

(1 + T S) l w

2 s

7 s

s 1

Where:

AT As defined in Note 1, t

I+753 As defined in Note 1, I+752 i

i 7

7 3,

2

- As defined in hote I, i

7 I

- As defined in Note 1, 1+7S 3

~

7

- -As defined in Note 1, 3

a As defined in Note 1, AT K.

1.09,

'K

- 0.02/oF' for increasing average temperature and 0 for decreasing average 3

{

temperature, 7S 7

The function generated by the rate-lag compensator' for T9 dynamic l+rS 7

compensation,'

.z o

i 4

w r,

Time constants utilized in the rate-lag coggensator for Tavg, r, - 10 s, w

I

~g As defined in Note 1, 1 + r.S.

t As defined in Note 1, 7

1

f f-

o.E o:

= ;:

TABLE 2.2-1 (Continued)

O gg TABLE NOTATIONS (Continued) g NOTE 3:

(Continued)

((

K 0.00180/*F for T > T" and K6 - O for T s T",

6 i

T As defined in Note 1, T"

Indicated T a

instrumentatV8n,t RATED THERMAL POWER (Calibration temperature for AT s ss7.ior),

3 S

As defined in Note 1, and f (AI)

O for all AI.-

4 2

7 IG UOTE 4:'

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

'2.7% AT span.

I i

4 8t a

o i

l i

.O 1

5

4m 2.1 SAFETY llMITS BASES 2.1.1 REACTOR CORE t

The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant.

Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the 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 the location of DNB for axially uniform and nonuni form heat flux distributions.

The local DNB heat flux ratio (DNBR) 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.

ncertainties in the WRB-1 or WRB-2 correlations, plant operating parameters, nuclear and thermal parameters, fuel fabrication parameters, and computer codes are considered statistically such that there is at least a 95 percent probability with 95 percent confidence level that DNBR will not occur on the most limiting fuel rod during Condition i and II events. This establishes a design DNBR value which must be met in piant safety analyses using values of input parameters without uncertainties.

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

The curves of Figures 2.1-1 and 2.1-2 show the loci of points of THERMAL POWER, Reactor Coolant System pressure, and average temperature below which the calculated DNBR is. no less than the design DNBR value or the average I

enthalpy at the vessel exit is less than the enthalpy of saturated liquid.

N These curves are based on an enthalpy hot channel factor, F H of 1.70 (includes measurement uncertainty) and a reference cosine with a da,k oful.55 for axial power shape.

An-allowance is included for an increase in F{H at reduced power based on the expression:

F H = 1.70 [1 + 0.3 (1-P)]

where P is the fraction of RATED THERMAL POWER These limiting heat flux conditions are highor than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion psuming axial imbalance is within the limits of F function of the Overtemperature trip.

When the axial power imbaldnc(e is not delta I) within the tolerance, the axial power imbalance effect on the Overtemperature delta T trips will reduce the setpoints to provide protection consistent with core safety limits.

MILLSTONE - UNIT 3 B 2-1 Amendment No. 60 0004

- _ _. ~ _,

LIMITING SAFETY SYSTEM SETTINGS BASES Intermediate and Source Ranae. Neutron Flux The Intermediate and Source Range, Neutron Flux trips provide core protection 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-of the Power Range, Neutron Flux channels.

The Source Range channels will initiate a Reactor trip at about 105 counts per second unless manually blocked when P 6 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 blocked when P-10 becomes active.

No credit was taken for operation of the trips associated with either the Intermediate or Source Range Channels-in the accident analyses; however, their functional capability at the specified trip settings is required by this specification to enhance the overall reliability of the Reactor Trip System.

Overtemoerature AT

"- Overtemperature AT tri all i,.umbinations of pressure, p provides core protection to prevent DNB for pwer, coolant temperature, and axial power distribution, provided that-the transient is slow with respect to piping transit delays from the core to the temperature detectors, and pressure is within the range between~ the Pressurizer High 'and -Low Pressure trips.

The-Setpoint 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 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 cifference between top and-bottom power range nuclear detectors, the Reactor-trip is automatically reduced according to the notations in Table 2.2-1.

Operation with 'a reactor coolant loop out of service--requires Reactor-Trip System modification. Three loop operation is permissible after resetting the K1 input to the Overtemperature AT channels, reducing the Power Range Neutron Flux High setpoint to a value just above the three : loop maximum permissible power level, and resetting-the P-8 setpoint to its three ' loop value.

These modifications have been chosen so that, in three loop operation, each component of the Reactor Trip System performs its normal four loop.

function, prevents operation outside the safety limit curves, and prevents the DNBR from going below the design limit during normal ' operational and antici-pated transients.

Overoower AT l

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 MILLSTONE - UNIT 3 B 2-5 Amendment No. JJ,60 0005

l LIMITING SAFETY SYSTEM SETTINGS BASES Steam Generator Water level The Steam Generator W,ter Level Low Low trip protects the reactor from loss of heat sink in the event of a sustained steam /feedwater flow mismatch resulting from loss of normal feedwater.

The specified Setpoint provides allowances for starting delays cf the Auxiliary Feedwater System.

Low Shaft Soeed Reactor Coolant Pumn The Low Shaf t Speed - Reactor Coolant Pumps trip provides core protection j

to prevent DNB in the event of a sudden significant decrease in reactor coolant pump speed (with resulting decrease in flow) on two reactor coolant pumps in any two operating reactor coolant loops.

The trip setpoint ensures that a reactor trip will be generated, considering instrument errors and response times, in sufficient time to allow the DNBR to be maintained greater than the design limit following a four-pump loss of flow event.

Turbine Trio A Turbine trip initiates a Reactor trip.

On decreasing power the Pactor trip from the Turbine trip is automatically blocked by P-9 (a power level of approximately 50% of RATED THERMAL POWER); and on increasing power, reinstated automatically by P 9.

Safety In.iection Inout from ESF If a Reactor trip has not already been generated by-the Reactor Trip System instrumentation, the ESF automatic actuation logic channels will initiate a Reactor trip upon any signal which initiates a Safety Injection.

The ESF instrumentation channels which ' initiate a Safety -Injection signal are shown in Table 3.3-3.

&uctor Trio System Interlocks The Reactor Trip System interlocks perform the following functions:

P6 On increasing power P-6 allows the manual block of the Source Range trip (i.e.,

prevents premature block of Source Range trip) and deenergizes the high voltage to-the detectors. On decreasing pos:er, Source Range Level trips are automatically reactivated and high vol-tage restored.

1 P-7 On increasing power P-7 automatically enables Reactor trips on low flow in more than one reactor coolant loop, reactor coolant pump low shaf t speed, pressurizer low pressure and pressurizer hi decreasing power, the above listed trips are automa' "gh level. On ly blocked.

MILLSTONE - UNITE 3 B 2-7 Amendment No.60 0006

,e l

3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1.1 BORATION CONTROL SHUT 00WN MARGIN - MODES 1 AND 2 LIMITING CONDITION FOR OPERATIAN 3.1.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to 1.3% Ak/k for both four loop and three 1c 9perat!on.

APPLICABILITY:

MODES I and 2*.

ACTION:

With the SHUTDOWN MARGIN les: than 1.3% Ak/k, immediately initiate and con-tinue boration at greater than or equal to 33 gpm of a solution containing greater than or equal to 6300 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.

SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1.3% Ak/k:

Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after detection of an inoperable control rod (s) and at a.

least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the rod (s) is inoperable, if the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod (s);

b.

When in MODE 1 or MODE 2 with Keff greater than or. equal to 1 at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that control bank withdrawal is within the limits of Specification 3.1.3.6; When in MODE 2 with K,ff less than 1, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to-c.

achieving reactor criticality by verifying that the predicted critical control rod position is within the -limits of Specification 3.1.3.6; d.

Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of Specification 4.1.1.1.2, with the control banks at the maximum insertion limit of Specification 3.1.3.6; and

See Special Test Exceptions Specification 3.10.1.

MILLSTONE --UNIT 3 3/4 1-1 Amendment No. 60 0007

REA:TIVITY CONTROL SYSTEMS SURVEILLANCE RE0VIREMENTS (Continued) 4.1.1.1.2 The overall-core reactivity balance shall be compared to predicted values to demonstrate agreement within 1% Ak/k at least once per 31 Effec-tive Full Power Days (EFPD). This comparison shall consider at least the following factors:

1)

Reactor Coolant System boron concentration, 2)

Control rod positian, 3)

Reactor Coolant System average temperature, 4)

Fuel burnup based on gross thermal energy generation, 5)

Xenon concentration, and 6)

Samarium concentration.

The predicted reactivity values shall be adjusted (normalized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 EFPD after each fuel loading.

l l

f MILLSTONE - UNIT 3 3/4 1-2 Amendment No. 60 0007

1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 B0 RATION CONTROL SHUTDOWN MARGIN - MODES 3. 4 AND 5 LOOPS FILLED LIMITING CONDITION FOR OPERATION 3.1.1.1.2 The SHUTDOWN MARGIN shall be greater than or equal to the limits shown in Figures 3.1-1, 3.1-3 and 3.1-4 for four loop operation and in Figure 3.1-2 for three loop operation.

APPLICABILITY:

MODES 3, 4 and 5 ACTION:

With the SHUTDOWN MARGIN less than the required value, immediately initiate and continue boration at greater than or equal to 33 gpm of a solution containing greater than or equal to 6300 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.

SURVEILLANCE RE0VIREMENTS 4.1.1.1.2.1 The SHUTDOWN MARGIN shall be determined to-be greater than or equal to the required value:

Within I hour after detection of an inoperable control rod (s) and at a.

least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the rod (s) is inoperable.

.If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod (s); and b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by consideration of the following factors:

1)

Reactor Coolant System boron concentration, 2)

Control rod position, 3)

Reactor Coolant System average temperature, 4)

Fuel burnup based on gross thermal energy generation, 5)

Xenon concentration, and 6)

Samarium concentration.

4.1.1.1.2.2 Valve 3CHS-V305 shall be verified closed and locked at least once per 31 days.

MILLSTONE - UNIT 3 3/4 1-3 Amendment No. 60 0007

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l WILLST0E - LAUT 3 -

3/414 Amendment No. 12,29,59,60 a.,

_.. _ _ _ ___ _ _ ____ _ _ _ _ __=._.,__ ___._,

.._,...,_......._.,-....,.,_,...,.....,,,,..,__m,.,_,,,,.,,,,,

4--

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(748,1.3001

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400 800 1200 1600 2000 2400 2900 g

RCS CRITICAL BOR0ff CODICE8tTRAT100f (ppel FIGURE 11-2 REE3JfRED 58tfTDOWN MARGM F(Nt RIODE 3 WITH TistEE LOOPS 195 (FERATIENI a

c E

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C2375,5.722)

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O 400 800 1200 1600 2000 2400 m

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RCS CRITICAL BORON CONCENTRATION (ppm)

FIGURE 3.1-3 REQUIRED SMJTDOWN MARCIN FOR M00E 4 i

E t"-G k

8.00 I

i c5

-4 i

7. x w

(2394. 5.964) 6.00 f

(1606, 5.1311

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1 2.x if 2-g I.00 (458,1.300).

3 n

=

0.00 8

0 400 000 1200 1600 2000 2400 2900 RCS CRITICAL BORON CONCO(TRATION (ppel l

FIGURE 3.1-4 l

REQUIRED SHUTDOWN MARGIN FOR UDDE 5 WITH RCS LOOPS FILLED

m REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN - COLD SHUTDOWN - LOOPS NOT FlLLED LIMITING CONDITION FOR OPERATION 3.1.1.2 The SHUTDOWN MARGIN shall be greater than or equal to a) the limits shown in Figure 3.1-5 or b) the limits shown in Figure 3.1-4 and secure the valves shown in Specification 4.4.1.4.2.3.

APPLICABILITY: MODE 5 LOOPS NOT FILLED ACTION:

With the SHUTDOWN MARGIN less than the above, immediately initiate and continue boration at greater than or equal to 33 gpm of a solution containing greater than or equal to 6300 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.

SURVEILLANCE REQUIREMENTS 4.1.1.2.1 The SHUTDOWN MARG!N shall be determined to be greater than or equal to the above:

Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months afte^ detection of an inoperable control rod (s) and at a.

least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the rod (s) is inoperable.

If the inoperable control rod is immovable or untrippable,'the SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod (s); and b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by consideration of the following factors:

1)

Reactor Coolant System boron concentration,.

2)

Control rod position, 3)

Reactor Coolant System average temperature, 4)

Fuel burnup based on gross thermal energy generation, 5)

Xenon concentration, and 6)

Samarium concentration.

4.1.1.2.2 Valve 3CHS-V305 shall be verified closed and lock =ed at least once per 31 days.

MILLSTONE - UNIT 3 3/4 1-8 Amendment No. 60 0007

e z

~

r-E

-4 K

10.00 o

k

-e 9.00 g

8.00 7.00

/-

=

k 3

(2356. 5.930) 6.00 (1395, 5.4943 5

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3.00 2.00

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093,1.300) g 0.00

~

0 400 800 1200 1600 2000 2400 2800 RCS CRITICAL BORON CONCDfTRATION (ppm)

FIGURE 3.1-5 REQUIRED SHUTDOWN MARGIN FOR MODE 5 Fim T3 LOOPS DRADED

-+

^

m REACTIVITY CONTROL SYSTEMS t!QQ[RATOR TEMPERATURE COEFFICIENT LIMITING CORDJTION FOR OPERATION 3.1.1.3 The moderator temperature coefficient (MTC) shall be within the limits specified in the CORE OPERATING LIMITS REP 0 The maximum upper limit shall-be less positive than +0.5 x 10'gTS (COLR).

Ak/k/'F for all the rods withdrawn, beginning of cycle life (BOL),. condition for power levels up to 70% RATED THERMAL POWER with a linear ramp to 0 Ak/k/'F at 100% RATED THERMAL POWER.

APPLICABIllTY: BOL - MODES 1 and 2* only**.

End of Cycle life (E0L) Limit - MODES 1, 2, and 3 only**.

ACTION:

With the MTC more positive than the BOL limit of Specification a.

3.1.1.3 above, operation in MODES 1 and 2 may proceed provided:

1.

Control rod withdrawal-limits are established and maintained sufficient to restore the MTC to less positive than-the above limits within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in H0T 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 limit for the all rods withdrawn condition; and 3.

A Special Report is prepared and submitted to thc Commission,

- pursuant to Specificatior. 6.9.2, within-10 days, describing the value of the mersured MTC, the interim control rod withdrawal limits, and the predicted ayerage core burnep necessary for restoring the posit *ve MIC to within its ilmit for the all rods withdrawn condition.

b.

With the MTC more negative than the E0L limit specified in the COLR, he in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

With K,f7 greater than or equal to 1.

- See S'pecial Test Exceptions Specification 3.10.3.

MILLSTONE - UNIT 3 3/4 1-10 Amendment No. 60 0007

).

..s REACTIVITY CONTROL SYSTEMS SURVEILLANCE REOUIREMENTS 4.1.1.3 The MTC shall be determined to be within its limits during each fuel cycle as follows:

o a.

The MTC shall be measured and compared to the BOL limit of Specification 3.1.1.3, above, prior to initial operation above 5%

of RATED THERMAL POWER, after each fuel loading; and b.

The MTC shall be measured at any THERMAL POWER and compared to the 300 ppm surveillance limit specified in the COLR (all rods withdrawn, RATED THERMAL POWER condition) within 7 EFPD after reaching an equilibrium boron concentration of 300 ppm.

In the event this comparison indicates the MTC is more negative 'than the 300 ppm surveillance limit specified in the COLR, the MTC shall be remeasured, and compared to the E0L MTC limit specified in the COLR, at least once per 14 EFPD during the remainder of the fuel cycle.

MILLSTONE - UNIT 3 3/4 1-11 Amendment No 172,60 0007

REACTIVITY CONTROL SYSTEMS MINIMUM TEMPERATURE FOR CRITICALITY LIMITING CONDITION FOR OPERATION 3.1.1.4 shall be greater than or equal to 551'F.The Reactor Coolant System lowest opera i

APPLICABILITY:

MODES 1 and 2* **.

l ACTION:

With a Reactor Coolant System operating loop temperature (7,yg) less than l

551*F, restore T,yg to within its 1 Mit within 15 minutes or be in HOT STANDBY within the next 15 minutes.

SURVEILLANCE RE0VIREMENTS 4.1.1.4 The Reactor Coolant System temperature (T"V9) shall be determined to be greater than or equal to 551*F:

Within 15 minutes prior to achieving reactor criticality, r.r.

a.

l b.

At least once per 30 rainutes when the reactor is critical and the l

Reactor Coolant System T is less than 5610F with the T

-T avg ayg ref Deviation Alarm not reset.

l l

t

With K,7f greater than or equal to 1.

- Swe Special Test Exceptions Specification 3.10.3.

~

l I

l 1

MILLSTONE - UNIT 3 3/4 1-12 Amendment No. 72, 60 0007

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

g REACTIVITY CONTROL SYSTEMS 3/4.1.2 BORAT10N SYSTEMS FLOW PATH SHUTDOWN LIMITING CONDITION FOR OPERATISN 3.1.2.1 As a minimum, one of the following boron injection flow paths shall be OPERABLE and capable of being ponered from an OPERABLE emergency power source:

a.

A flow path from the boric acid storage system via either a boric acid transfer pump or a gravity feed connection and a charging pump to the Reactor Coolant System if the boric acid storage system in Specification 3.1.2.5a. is OPERABLE, or b.

The flow sth from the refueling water storage tank via a charging pump to tie Reactor Coolant System if the refueling water storage tank in Specification 3.1.2.5b. is OPERABLE.

APPLICABit.lTY: MODES S and 6.

ACTION:

With none of the above flow paths OPERABLE or capable of being powered from an OPERABLE emergency power source, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

SURVEILLANCE REQUIREMENTS 4.1.2.1 At least one of the above required flow paths shall be demonstrated OPERABLE:

a, At least once per 7 days by verifying that the Boric Acid Transfer Pump Room temperature and the boric acid storage tank solution temperature are greater than or equal to 678F when a flow path from the boric acid tanks is used, and b.

At least once per 31 days by verifying tl.at each valve (manual, power-operated, or automatic) in the flow path that is not locked, i

sealed, or otherwise secured in position, is in its correct j

position, l

MILLSTONE - UNIT 3 3/4 1-13 Amendment.No. 60 l

0001

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

REACTIVITY CONTROL SYSTEMS FLOW PATHS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.2 At least two* of the following three boron injection flow paths shall be OPERABLE:

a.

The flow path from the boric acid storage system via a boric acid transfer pump and a charging pump to the Reactor Coolant System (RCS),and b.

Two flow paths from the refuelina water storage tank via charging pumps to the RCS.

APPLICABILITY:

MODES 1, 2, 3, and 4.

EllDti:

With only one of the above n quired boron injection flow paths to the RCS OPERABLE, restore at least two boron injection flow paths to the RCS to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY and borated to a SHUTDOWN MRGIN equivalent to at least the limits as shown in Figure 3.14 at 200'F within the next 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months st restore at least two flow paths to OPERABLE status within the next 7 days or be in COLD SHVTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

M RVEILLANCE RE0VIREMENTS 4.1.2.2 At least two of the above required flow paths shw#1 be demonstrated OPERABLE:

a.

At least once per 7 days by verifying that the Boric Acid Transfer Pump Room temperature and the boric acid storage tank solution temperature are greater than or equal to 67'F when it is a required water source; b.

At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct

position, At least cnce per 18 months during shutdown by verifying that each c.

automatic valve in the flow path actuates to its correct position on a Safety injection test signal; and d.

At least once per 18 months by verifying that the flow path required by Specification 3.1.2.2a. delivers at least 33 gpm to i

the RCS.

0nly one boron injection flow path is required to be OPERABLE whenever the temperature of one or more of the RCS cold legs is less than or equal to 350*F.

MILLSTONE UNIT 3 3/4 1-14 Amendment No. 59,60 0007

i r

REACTIVITY CONTROL SYSTEMS i

(lWtGING PUMP SHUTDOWN LIMITING CONDITION FOR OPERATION I

4 i

3.1.2.3 One charging pump in the boron injection flow path required by

{

Specification 3.1.2.1 shall be OPERABLE and capable of being powered from an j

OPERABLE emergency power source.

APPLICABILITY.: MODES 5 and 6.

4 ACTION:

[

With no charging pump OPERABLE or capable of being powered from an OPEFABLE i

emergency power source, suspend all operations involving CORE ALTERATIONS or i

positive reactivity changes.

4 SURVEllLANCE REOUIREMENTS l

~

4.1.2.3.1 The above required chargin by verifying, on recirculation flow, g pump shall be demonstrated OPERABLE that a sifferential pressure across the t

pump of greater than or equal to 2411 psid is developed when tested pursuant to Specification 4.0.5.

4.1.2.3.2 All charging pumps, excluding the above required OPERABLE pump, shall be demonstrated inoperable at least once per 31 days, except when the reactor vessel head is removed, by verifying that the motor circuit breakers i

are secured in the opei) position.

^

l t

i i

i l

i' MILLSTONE - UNIT 3 3/4 1-15 Amendment No.'50,60 0007

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_ _. _ _. _ _ _ _ _. ~. _

REACTIVITY CONTROL SYSTEMS CB SGING PUMPS OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.4 At least two* charging pumps shall be OPERABLE.

i APPLICABILITY: MODE 5 1, 2, 3, and 4.

ACTION:

With only one charging pum) OPERABLE, restore at least two charging pumps to i

OPERABLE status within 72 aours or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least the limit as shown in Figure 3.1 4 at 200'F withis the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore at least two charging pumps to OPERABLE status within the next 7 days or be in COLD SHUTDDWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE RE0VIREMENTS 4.1.2.4.1 At least two charging pumps shall be demonstrated OPERABLE by verifying, on recirculation flow, that a differential pressure across each i

pump of greater than or equal to 2411 psid is developed when tested pursuant to Specification 4.0.5.

4.1.2.4.2 All charging pumps, except the above allowed OPERABLE pump, shall be demonstrated inoperable at least once per 31 days whenever the l

temperature of one or more of the Reactor Coolant System (RCS) cold legs is l

less than or equal to 350'F by verifying that the motor circuit breakers are secured in the open position, i

A maximum of one centrifugal charging pump shall be OPERABLE whenever the temperature of one or more of the RCS cold legs is less than or equal to

{

350'F.

MILLSTONE - UNIT 3 3/4 1-16 Amendment No. 60 0007

.. ~..., _

1 REACTIVliY CONTROL SYSTEMS BORATED WATER SOURCE SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.5 As a minimum, one of the following borated water teurces shall be OPERABLE:

a.

A Boric Acid Storage System with:

1)

A minimum contained borated water volume of 6700 gallons, 1

2)

A boron concentration between 6300 and 7175 ppm, and 3)

A minimum solution temperature of 67'F.

b.

The refueling water storage tank (RWST) with:-

1)

A minimum contained borated water volume of 250,000 gallons,.

2)

A minimum boron concentration of 2700 ppm, and 3)

A minimum solution temperature of 40'F.

APPLICABILITY: MODES 5 and 6.

ACTION:

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

SURVEILLANCE RE00lREMENTS 4.1.2.5 The above required borated water source shall be demonstrated OPERABLE:

a.

At least once per 7 days by:

1)

Verifying the boron concentration of tho water, 2)

Verifying the contained borated water volume, and 3)

Verifying the Boric Acid Transfer Pump,'oom temperature and the boric acid storage tank solution temperature when it is-the source of borated water.

b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature when it is the source of borated-water and the outside air temperature is less than 35'F.

MILLSTONE - UNIT 3 3/4 1-17 Amendment No.60 0007

,_,..._,._.-.__..___,_.,_-__,_....,,_[.

j a,.

i REACTIVITY CONTROL SYSTEM BORATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION t

3.1.2.6 As a minimum the following borated water source'(s) shall be OPERABLE as required by Specification 3.1.2.2:

a.

A Boric Acid Storage System with:

1)

A minimum borated water usable volume of 21,020 gallons, f

2)

A boron concentration between 6300 and 7175 ppm, and 3)-

A minimum solution temperature of 67'F.

b.

The refueling water storage tank (RWST) with:

1)

A minimum contained borated water volume of 1,166,000

gallons, ji 2)

A boron concentration between 2700 and 2900 ppm,

(

3)

A minimum solution temperature of 40'F, and f

4)

A maximum solution temperature of 50*F.

~

APPLICABILITY: MODES 1, 2, 3, and 4.

MJ.10.ti:

l With the Boric Acid Storage System inoserable, restore the system a.

i to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or >e in at least HOT STANDBY.

i within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and borated to a SHUTDOWN MARGIN equivalent to at least the limits as shown in Figure 3.1-4 at 200'F; restore the Boric Acid Storage System to OPERABLE status within the next 7 days or be_in COLD Sh0TDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b.

With the RWST inoperable, restore the tank to OPERABLE status within I hour or be in at least HOT STANDBY within the next t

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

i I

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L l

MILLSTONE - UNIT 3' 3/4 1-18 Amendment No. 60 0007 i

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I REACTIVITY CONTROL SYSTEMS SEVEILLANCE REQUIREMENTS 4.1.2.6 Each borated water source shall be demonstrated OPERABLE:

a.

At least once per 7 days by:

1)

Verifying the boron concentration in the water, 2)

Verifying the contained borated water volume of the water source, and 3)

Verifying the Boric Acid Transfer Pump Room temperature and the boric acid storage tank solution temperature.

b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature, i

MILLSTONE - UNIT 3 3/4 1 19 Amendment No. 60 0007

_ _ ~ _. _. _ _ _ _ _. __ _ _ _. _ _ _ _

'd-:p REACTIVITY CONTROL SYST1M1 3/4.1.3 MOVABLE CONTROL ASSEMBLIES GROUP HEIGHT LIMITING CONDITION FOR OPERATION I

i 3.1.3.1 All full length shutdown and control rods shall be OPERABLE and positioned within 112 steps (indicated position) of their group step counter l

demand position.

APPLICABILITY:

MODES 1* and 2*.

ACTION:

a.

With one or more full-length rods inoperable due to being immovable as a result of excessive friction or mechanical interference or known to be untrippable, determine that the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied within I hour and be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , b.

With one full length rod trippable but inoperable due to causes other than addressed by ACTION a., above, or misaligned from its group step counter demand height by more than 112 steps (indicated position), POWER OPERATION may continue provided that within I hour:

l 1.

The rod is restored to OPERABLE status within the above alignment requirements, or 2.

The rod is declared inoperable and the remainder of the rods in the group with the inoperable rod are aligned to within i

12 steps of the inoperable rod while maintaining the rod sequence and in.ertion limits of Specification 3.1.3.6.

The THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation, or 3.

The rod is declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.1-is satisfied.

POWER OPERATION may then continue provided that:

a)

A reevaluation of each accident analysis of Table 3.1-1 I

is performed within 5 days; this reevaluation shall confirm that the previously analyzed results of these accidents remain valid for the duration of operation under these conditions; b)

The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ;

See Special Test Exceptions Specifications 3.10.2 and 3.10.3.

MILLSTONE - UNIT 3 3/4 1 20 Amendment No. JS.60 0001

w REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION MJJON (Continued) c)

Apowerdistributionmapisobtagnedfromthemovable incore detectors and F (Z) and I are verified to be within their limits wikhin 72 hobYst and d)

With four loops o)erating, the THERMAL POWER level is reduced to less t1an or equal to M% of RATED THERMAL POWER within the next hour and within the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the High Neutron Flux Trip Setpoint is reduced to less than or equal to 85% of RATED THERMAL POWER, or e)

With three loops operating, the THERMAL POWER level is reduced to less than or equal to 50% of RATED THERMAL POWER within the next hour and within the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the Neutron Flux High Trip Setpoint is reduced to less than or equal to 60% of RATED THERMAL POWER.

c.

With more than one rod trippable but inoperable due to causes other than addressed by ACTION a above, POWER OPERATION may continue provided thatt 1.

Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , the remainder of the rods in the bank (s) with the ino)erable rods are aligned to within 112 steps of the inopera)1e rods while maintaining the rod sequence and insertion limits of Specification 3.1.3.6.

The THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation, and 2.

The inoperable rods are restored to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , d.

With more than one rod misaligned from its group step counter demand height by more than 112 steps (indicated position), be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , l

SURVEILLANCE RE0VIREMENTS 4.1.3.1.1 The position of each full-length rod shall be determined to be within the group demand limit by verifying the individual rod positions at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the rod position deviation monitor is inoperable, then verify the group positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

4.1.3.1.2 Each. full-length rod not fully inserted in the core shall be determired to be OPERABLE by movement of at least 10 steps in any one oirection at least once per 31 days.

1 MILLSTONE - UNIT 3 3/4 1 21 Amendment No. #,60 0007

?

i i

TABLE 3.1 1 i

l ACCIDENT ANALYSES REQUIRING REEVALUATIQN l

IN THE EVENT OF AN INOPERABLE FULL-LENGTH ROD Rod Cluster Control Assembly Insertion Characteristics Rod Cluster Control Assembly Misalignment Loss of Reactor Coolant from.Small Ruptured Pipes or from Cracks in large Pipes Which Actuates the Emergency Core Cooling System i

Single Rod Cluster Control Assembly Withdrawal at Full Power Major Reactor Coolant System Pipe Ruptures (Loss of Coolant Accident) l Major Secondary Coolant System Pipe Rupture Rupture of a Control Rod Drive Mechanism Housing (Rod Cluster Control Assembly Ejection)

L l

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i l

l l

l i

i 1

i l

MILLSTONE - UNIT 3 3/41-22 Amendment No, f>p,60-0007-

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

. ~

REACTIVITY CONTROL SYSJ E POSITION INDICATION SYSTEMS - OPERATING 1

LIMITING CONDITION FOR ODERATION 3.1.3.2 The Digital Rod Position Indication System and the Demand Position Indication System shall be OPERABLE and capable of determining the control rod positions within t u steps.

APPLICABIL111: MODES 1 and 2.

ACTION:

With a maximum of one digital rod position indicator per bank a.

inoperable:

1.

Determinethepositionofthenonindicatingrod(s) indirectly by the movable incore detectors at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and imediately after any motion of the nonindicating rod which exceeds 24 steps in one direction since the last dotednation of the rod's position, or 2.

With four loops operating, reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , or 3.

With three loops operating, reduce THERMAL POWER to less than 32% of RATED THERMAL POWER within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , b.

With a maximum of one demand position indicator per bank inoperable:

1.

Verify that all digital rod position indicators for the affected bank are OPERABLE and that the most withdrawn rod and the least withdrawn rod of the bank are within a maximum of 12 steps of each other at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , or 2.

With four loops operating, reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , or 3.

With threa loops operating, reduce THERMAL POWER to less than 32% of RATED THERMAL POWER within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

(

SURVEILLANCE RE0VIREMENTS l

-4.1.3.2 Each digital rod position indicator shall be determined to be OPERABLE by verifying that the Demand Position Indication System and the Digital Rod Position Indication System agree within 12 steps at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the rod position deviation monitor is inoperable, then compare the Demand Position Indication System and the Digital Rod Position Indication System at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

l MILLSTONE - UNIT 3 3/4 1-23 Amendment No. E9. 60 0001 I

.._=

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REACTIVITY CONTROL SYSTEMS POSITION INDICATION SYSTEM - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.3.3 One digital rod position indicator excluding demand position indication) shall be OPERABLE and caaable of(determining the control rod position within 112 steps for each slutdown or control rod not fully inserted.

APPLICABillTY: MODES ?.* **, 4* **, and 5* **.

ACTION:

Withlessthantheaboverequiredpositionindicator(s) OPERABLE, immediately open the Reactor Trip System breakers.

SURVEILLANCE RE0VIREMENTS 4.1.3.3 Each of the above required digital rod sosition indicator be determined to be OPERABLE by verifying that tie digital rod pos(s) shall ition indicators agree with the demand position indicators within 12 steps when exercised over the full-range of rod travel at least once per 18 months.

~ With the Reactor Trip System breakers in the closed position.

l

- See Special Test Exceptions Specification 3.10.5.

l i

MILLSTONE - UNIT 3 3/4 1-24 Amendment No. 60 ocer a

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REACTIVITY CONTROL SYSTEMS f

R00 DROP TIME 4

LIMITING CONDITION FOR OPERATION 3.1.3.4 The individual full length (shutdown and cont M1 r'od drop time from the fully withdrawn position shall be less than or eq)'ual to 2.7 seconds i

from beginning of decay of stationary gripper coil voltage to dashpot entry with:

3 T,yg greater than or equal to 5518F, and a.

\\

b.

All reactor coolant pumps operating.

APPLICABILITY: MODES I and 2.

ACTION:

a.

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 i

prior to proceeding to MODE 1 or 2.

t b.

With the rod drop timos within limits but determined with three reactor coolant pumps operating, operation may proceed provided THERMAL POWER is restricted to less than or equal to 65%_of RATED THERMAL POWER with the reactor coolant stop valves in the nonoperating loop closed.

+

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

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

b.-

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

At least once per 18 months.

It

(

k MILLSTONE - UNIT 3 3/4 1-25 Amendment No. 60-0001 l

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REACTIVITY CONTROL SYSTEMS SHUTDOWN ROD INSERTION LIMil LIMITING CONDITION FOR OPERATION 3,1.3.5 All shutdown rods shall be limited in physical insertion as specified in the core operating limits report (COLR).

APPLICABILITY: MODES 1* and 2* **.

ACTION:

With a maximum of one shutdown rod inserted beyond the insertion limits specified in the COLR except for surveillance testing pursuant to Specification 4.1.3.1.2, within I hour either:

Restore the rod to within the limit specified in the COLR, or a.

E.,

Declare the rod to be inoperable and apply Specification 3.1.3.1.

SURVEILLANCE RE0VIREMENTS 4.1.3.5 Each shutdown rod shall be determined to be within the insertion limits specified in the COLR:

Within 15 minutes prior to withdrawal of any rods in Control Bank a.

A, B, C, or D during an approach to reactor criticality, and b.

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

See Special Test Exceptions Specifications 3.10.2 and 3.10.3.

- With K,ff greater than or equal to 1.

1 l

MILLSTONE -UNIT 3 3/4 1-26 Amendment No. 60 l

ooor l

- - - - - - - - - - - - - - - - - - - - ~ - ~

REACTIVITY CONTROL SYSTEMS CONTROL ROD INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.6 The control banks shall be limited in physical insertion as specified in the core operating limits report (COLR).

APPLICABILITY: MODES 1* and 2* **.

ACTION:

With the control banks inserted beyond the insertion limits specified in the COLR, except for surveillance testing pursuant to Specification 4.1.3.1.2:

4 Restore the control banks to within the limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or a.

b.

Reduce THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the bank posi-tion using the insertion limits specified in the COLR, or c.

Be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REOUIREMENTS 4.1.3.6 The position of each control bank shall be determined to be within the insertion limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the rod insertion limit monitor is inoperable, then verify the individual rod positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

See Special Test Exceptions Specifications 3.10.2 and 3.10.3.

- With K,ff greater than or equal to 1.

l i

MILLSTONE - UNIT 3 3/4 1 27 Amendment No. J#.60 -

0001 J

2/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE FOUR LOOPS OPERATING LIMITING CONDITION FOR OPERATION 3.2.1.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within:

The limits specified in the CORE OPERATING LlHITS REPORT (COLR) for a.

Relaxed Axial Offset Control (RA00) operation, or b.

Within the target band about the target flux difference during base load operation, specified in the COLR.

APPLICABIllTY: MODE I above 50% RATED THERMAL POWER *.

EJIM:

For RAOC operation with the indicated AFD outside of the applicable a.

limits specified in the COLR, 1.

Either restore the indicated AFD to within the COLR specified limits within 15 minutes, or 2.

Reduce THEPJ'AL POWER to less than 50% of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux High Trip 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 , b.

For base load operation above APLND with the indicted AFD outside of the applicable target band about the target flux differences:

1.

Either restore the indicated AFD to within the COLR specified l

target band within 15 minutes, or 2.

Reduce THERMAL POWER to less than APLND of RATED THERMAL POWER and discontinue base load operation within 30 minutes, THERMAL POWER shall not be increased above 50% of RATED THERMAL c.

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

See Special Test Exception 3.10.2 MILLSTONE UNIT 3 3/4 2-1 Amendment No. g,60 0011

-~-

I f

i l

I POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS l

4.2.1.1.1 The indicated AFD shall be determined to be within its limits 1

l during POWER OPERATION above 50% of RATED THERMAL POWER by:

j 6

a.

Monitoring the indicated AFD for each OPERABLE excore channel at t

i least once per 7 days when the AFD Monitor Alarm is OPERABLE:

b.

Monitorin and logging the indicated AFD for each OPERABLE excore channel a 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 ti.e AFD Monitor Alarm is inoperable.

The logged values of the indcated AFD shall be assumed to exist during the interval preceding each logging.

l 4.2.1.1.2 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 l

limits.

l 4.2.1.1.3 When in base load operation, the target flux difference of each OPERABLE excore channel shall be determined by measurement at least once per 92 Effective Full Power Days.

The provisions of Specification 4.0.4 are not i

applicable.

4.2.1.1.4 When in base load operation, the target flux difference shall be i

updated at least once per 31 Effective Full Power Days by either determining the target flux difference in conjunction with the surveillance requirements of Specification 4.2.1.1.3 or by linear interpolation between the most recently measured value and the calculated value at the end of cycle life.

The provisions of Specification 4.0.4 are not applicable.

6 e

i i

i MILLSTONE - UNIT 3 3/4 2 2 Amendment No. M,60.

0011 F

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

,m

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

. ~.. - -.. _.

f POWER DISTRIBUTION llMITS AXIAL FLULDIFFERENCE

.THREE LOOPS OPERATING LIMITING CONDITION FOR OPERATION 4

3.2.1.2 The indicated AX1AL FLUX DIFFERENCE (AFD) shall be maintained within:

The limits specified in the CORE OPERATING LIMITS REPORT (COLR) for a.

Relaxed Axial Offset Control (RA0C) operation, or b.

Within the target band specified in the COLR about the target flux difference during base load operation.

M.PLICABillTY: MODE I above 37.5% of RATED THERMAL POWER.*

ACTION:

For RAOC operation with the indicated AfD outside of the applicable a.

limits specified in the COLR, 1.

Either restore the indicated AFD to within the COLR specified limits within 15 minutes, or 2.

Reduce THERMAL POWER to less than 37.5% of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux--High Trip-setpoints to less than or equal to 41% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , b.

For base load operation above APLND with the indicated AFD outside of the applicable target band about the target flux differences:

1.

Either restore the indicated AFD to within the COLR specified i

target band within 15 minutes, or 2.

Reduce THERMAL POWER to less than APLND of RATED THERMAL POWER and discontinue base load operation within 30 minutes.

THERMAL POWER shall not be increased above 37.5% of RATED THERKAL c.

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

See Special Test Exception 3.10.2.

MILLSTONE - UNIT 3 3/4 2 3 Amendment No.. g,60 l

0011

.~

POWER DISTRIBUTIQN LIMITS SURVEILLANCE RE0VIREMENTS 4.2.1.2.1 The indicated AFD shall be determined to b'e within its limits during POWER OPERATION above 37.5% 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:

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.

4.2.1.2.2 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.

4.2.1.2.3 When in base load operation, the target flux difference of each OPERABLE excore channel shall be determined by measurement at least once per 92 Effective Full Power Days.

The provisions of Specification 4.0.4 are not applicable.

4.2.1.2.4 When in base load operation, the target flux difference shall be updated at least once per 31 Effective full Power Days by either determining the target flux difference in conjunction with the surveillance requirements of Specification 4.2.1.2.3 or by li w r interpolation between the most recently measured value and the cablated value at the end of cycle life.

The provisions of Specification 4.0.4 are not applicable.

f l

l MILLSTONE - UNIT 3 3/4 2-4 Amendment No S.60 0011

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POWER DISTRIBUTION LIMITS

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3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - E 1Z.),

g FOUR LOOPS OPERATING LIMITING CONDITION FOR OPERATION i

3.2.2.1 F (Z) shall be limited by the following relationships:

g F(Z)sFhTPK(Z)forP>0.5 g

P RTP F(Z)sF K(Z) for P s 0.5 g

O 0.5 RTP F

the F limit at RATED THERMAL POWER (RTP) provided in theco=reoperStinglimitsreport(COLR).

Where:

P

THERMAL POWER

, and RATED THERMAL POWER h

K(Z) = the normalized F(Z) as a function of core height i

specified in the COLR.

g I

APPLICABILITY: MODE 1.

.i ACTION:

i With F (Z) exceeding its limit:

?

n I

Reduce THERMAL POWER at least 1% for each 1% fn(Z) exceeds the a.

limit within 15 minutes 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 ; POWER j

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 reduced at least 1% for each 1% F (Z) exceeds the limit, and g

I b.

Identify and correct the cause of the out of-limit condition l

prior to increasing THERMAL POWER above the reduced limit re-

{

quired by ACTION a.,-above; THERMAL POWER may then be increased i

provided Fn(Z) is demonstrated through incore mapping to _be i

within its ' limit.

MILLSTONE - UNIT 3 3/4 2 5

' Amendment No. Ap,60-0011

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

_.-.____.-.___.---__._-.__.__,,_.,._.-.__._,,__[.

i POWER DISTRIBUTION LIMITS SURVEILLANCE RE0VIREMENTS q

4.2.2.1.1 The provisions of Specification 4.0.4 are not applicable.

4.2.2.1.2 For RAOC operation, F (z) shall be evaluated to determine if F (z) g g

is within its limit by:

4 a.

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

b.

Increasing the measured F (z) component of the power distribution fok manufacturing tolerances and further map by 3% to account increasing the value by 5% to account for measurement uncertainties.

Verify the requirements of Specification 3.2.2.1 are satisfied, c.

Satisfying the following relationship:

M F"

p (g) g 0 x K(z) for P > 0.5 P x W(z)

M F

p (z) s x K(z) for P s 0.5 W(z) x 0.5 where FN(z) is the measured Fn(z) increased by the agances fo manufact'uring tolerances and me'hsurement uncertainty, F is the F limit, K(z) is the normalized F (z) as a function of coke height, h 9

l 1s the relative THERMAL POWER, and W(z) is the cycle-dependent functionthataccountsfor_gerdistribution.transientsencountered during normal operation.

F l

CORE OPERATING LlHITS REPORf as, pe(r ) Specification 6.9.1.6.K d.

Measuring F (z) according to the following schedule:

(1) Upon achieving equilibrium conditions after exceeding by 10% or more of RATED THERMAL POWER, the THERMAL POWER at which F (z) was last determined,* or-9

During power escalation at the beginning-of each cycle, power level may be increased until a power level for extended operation has been achieved and power distribution map outlined.

l MILLSTONE - UNIT 3 3/4 2-6 Amendment _No Jp,60 0011

i v

1 i

POWER DISTRIBUTION LIMITS j

$URVEILLANCE REQUIREMENTS (Continued) i, (2) At least once per 31 Effective Full Power Days, whichever occurs first.

l e.

With the maximum value of F(z)

K(7) over the core height (z) increasing since the previous determination of F (z), either of the following actions shall be taken:

(1)

F (z) shall be increased by 2% over that specified in Specifi-c tion 4.2.2.1.2c, or (2) F (z) shall be measured at least once per 7 Effective Full P wer Days until two successive maps inoicate that the maximum value of F (z)

K(z) over the core height (z) is not increasing.

1 f.

With the relationships specified in Specification 4.2.2.1.2c not being satisfied:

i (1) Calculate the maximum percent over the core height (z) that F(z)exceedsitslimitbythefollowingexpression:

q

~f(z)xW(z)

F 1

x 100 for P 1 0.5 RTP 0

. x K(z)-

P l

l l

i I

MILLSTONE - UNIT 3 3/4 2-7

. Amendirent No. JJ,60 0011 "1E T'*-.9-@t*

1'r w-tar.-owrwm*---W.-9gy,w y

wm,-up-e,-weymyy-gwee--gesy,,.me.e.e,-yy,c ey rp.,%

y P

y y'

g--wypoe m999-.

y p.ggpm.-gy--,ycgry,ge*g -wa 9w ww owyg weg.c-ag'rT"9'" M Tw9 4 r W v v W D

1 i

1 POWER OlSTRIBUTION LIMITS SURVEILLANCE REOUIREMENTS (Continued) 1 I

F (z) x W(z)

RTP 1

x 100 for P < 0.5

_ L X K(z) 0.5 (2) One of the following actions shall be taken:

(a) Within 15 minutes, control the AFD to within new AFD limits which are deteimined by reducing the applicable AFD limits by 1% AFD for each percent Fn(z) exceeds its limits as determined in Spec 1fication 1.2.2.1.2f l.

Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , reset the AFD alarm setpoints to these modified limits, or (b) Comply with the requirements of Specification 3.2.2.1 for F (z) exceeding its limit by the percent calculated, or g

(c)

Verify that the requirements of Specification 4.2.2.1.3 for base load operation are satisfied and enter base load operation.

g.

The limits specified in Specifications 4.2.2.1.2c, 4.2.2.1.2e, and j

4.2.2.1.2f above are not applicable in the following core plane regions:

(1) Lower core region from 0% to 15%, inclusive.

(2) Upper core region from 85% to 100%, inclusive.

4.2.2.1.3 Base load operation is permitted at powers above APLND if the following conditions are satisfied:

Prig to entering base load operation, maintain THERMAL POWER above a.

APL and less than or equal to that allowed by Specifica-tion 4.2.2.1.2 for at least the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Maintain base-load operation surveillance (AFD within the target band limit the target flux difference of Specification 3.2.1.1) during this time period.

Base load operation g then pg[mitted providig THERMAL l

POWER is maintained between APL and APL or between APL and l

MILLSTONE - UNIT 3 3/42-8 Amendment No. JJ,60 0011

~ -, _.

rd i

POWER DISTRIBUTION LIMITS i

l SURVEILLANCE REQUIREMENTS (Continued) i 1005 (whichever is most limiting) and Fg sur{et11ance is maintained--

l pursuant to Specification 4.2.2.1.4.

APL is defined as the-g minimum value oft i

F xK(z)

RTP APLOL =

0 x 100%

4 MF(z)x-W(2)gg over the core height (z) where:

F' (z) is the measured F (z) o increased by the alloyances for manuf cgping tolerances and mea-surement uncertainty.

The Fn limit is F W(z) is the cycle.

n 1

dependent functir/ that ace' bunts for l'imited pow Qistribution transient encountered during base load operation.

F

, K(z), and W(z)gt arespecifiedintheCOLRasperSpecificationk.9.1.6.

b.

Durgg base load operation, if the THERMAL POWER is decreased below-APL then the conditions of 4.2.2.1.3.a shall-be satisfied before reentering base load operation.

4.2.7.1.4 During base load operation F (z) shall be evaluated to determine if F lz) is within its limit by:

g q

a.

Using tht movable incore detectorg to obtain a power distribution map'at any THERMAL POWER above APL I

b.

Increasing the measured F (z) component of the power distribution 0

map by 3% to account for manufacturing tolerances and further l

increasing the value by 5% to account for measurement uncertainties, Verify the requirements of-Specification 3.2.2.1 are satisfied, t

j c.

Satisfying the following relationship:

j F

x K(z) for P > APL p g)g 0

{

PxW(z)gt N

RTP where:

F z) is the measured F F

is the F limit, the normalizeh(Fn(z) as a function ok(z). core hSigh'..

P is khe relative i

j THERMAL POWER. W(z)gt is the cycle dependent function that accounts s

i

- MILLSTONE - UNIT 3 3/412 9

-Amendment No. p,60 0011-j

.. ~.

l

~

POWER DISTRIBUTION LIMITS f

SURVEILLANCE REOUIREMENTS (Continued) i j

l for limited power gstribution transients encountered during base R

load operation.

F

,-K(z),andW(z)BLarespecifiedintheCOLRas o

per Specification 6.9.1.6.

I N

d.

Measuring F (z) in con,iunction with target flux difference determi-n nation according to the following schedule:

i (1)

Prior to entering base load operation after satisfying Sec-tion 4.2.2.1.3 unless a full core flux map has been taken in the previous 31 EFPD witg the relative thermal power havitig i

been maintained above APL for the 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to mapping, l

and 4

(2) At least once per 31 Effective Full Power Days.

e.

With the maximum value of F (z)

K(z)

I j

over the core height (z) increasing since the previous determination of F (z), either of the following actions shall be takers:

(1)

F (z) shall be increased by 2% over that specified in 4.2.2.1.4.c, or l

(2)

F$(z) shall be measured at least once -per 7 Effective Full l

Power Days until 2 successive maps indicate that the maximum l

value of Ff(z)

I K(z) overthecoreheight(z)isnotincr<aasing.

l l

MILLSTONE - UNIT 3 3/4 2-10 Amendmeat No. JS.60 ocu

POWER DISTRIBUTION LINES R RVEILLANCE REQUllW WITS (Continued) f.

With the relationship specified in 4.2.2.1.4.c not being satisfied, either of the following actions shall be taken:

f (1)

Place core in an equilibrium conditior where the limit in 4.2.2.1.2.C is satisfied, and remeasure F' (z), or (2) Comply with the requirements of Specification 3.2.2.1 for F (t) g exceeding its limit by the maximum percent calculated over the core height (z) with the following expression:

i F$(z)xW(z)BI,

!!D 1

x 100 for P 2 APL RTP

{L. x K(z)

P d

g.

The limits specified in 4.2.2.1.4.c 4.2.2.1.4.e. and 4.2.2.1.4. f s

are not applicable in the following core plans regions:

(1)

Lower core region 0% to 150 inclusive.

(2) Upper core region 85% to 100%, inclusive.

4.2.2.1.5 When Fn(z) is measured for reasons other than meeting the require-ments of Specificlition 4.2.2.1.2, an overall measured F (z) shall be obtained 0

from a power distribution map and increased by 3% to account fw manufacturing tolerances and further increased by 5% to account for measureent unce:tainty.

i l

l l

MILLSTONE. UNil 3 3/4 2 11 Amendment No. A9,60 0011 s

-+,.f.m rv t

-r ---r-i-e ' e

+v-*-'

--.uenr, mw r-~*w+w-e--

,. *,w--

orv w

=w a-s-e

~w-~r-e-,--*=^-s--+es-swa-,e--*we 4=a e

ena~

s-i=w-ew&>*d-v=-e-we-m

-s-6w

POWER DISTRIBUTION 1.lMITS HEAT fVI' 40T t % Ntt FACTOR F (Z) 1 n

FREE' LOOPS OPERATING uY111NG CONDIT10N FOR OPERAT10N,. _.

3.2.2.2 F (Z) shall be limited by i.h following relationships:

n rRTP F (l:

'. (KI,Z)l

ir P > 0.375 q

RTP F(Z)s([0

) (K(Z)) for P 10.375 0

0.375 I

)

RTP F

The F Halt t RATED THERMAL POWER (RTP) specified in the CORE OPEbT.NG LIM (O REPORT (COLR).

(

s P

RATED THERMAL POWER

{

TBEGALB t :{t and-Where:

L K(Z) = the normalized F (2) as a function of core height speci-f fled in the COLR.

APPLICABILITX: MODE 1.

ACTION:

With F (Z) exceeding its lim L.

g a.

Reduca THERMA ( POWER at he ' 1% 5 each 1% Fn(Z) exceeds the limit within 15 minutes ar-f.iminuly-- reduce 'the Power Range 3

Neutron Flux High Trip Setpdah within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; POWER-OPERAUON may proceed for up sd totai of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; subsequent POWER WERATION may proceed proslod the Overpower AT Trip Set-points have been reduced at least L' for each 1% Fn(Z) exceeds the limit.

The Overpower AT Trip F tpoint reductfbn shall be porfora d with the reactor in at len HOT STANDBY.

b.

Identify and correct the cause of tt 3 out of limit coMition prior to increasing THERMAL F. *.R above the reduced lini. re-quired by M 710N a.,. above; THEnMAL POWER may then be inctvsed provided Fn(Z) is demonstrated.through incore mapping to M within its '11mit.

Millstone - Unit 3 3/4 2 Amendment No. #,60 0011 i

on~-

4 i

l POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 4.2.2.2.1 The provisions of Specification 3.0.4 are not applicable.

4.2.2.2.2 For RAOC operation, F (z) shall be evaluated to determine if F (z) is g

g within its limit by:

Using the movable incore detectors 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 map 0

by 3% to account for manGfacturing tolerances and further increasing the value by 5% to account for measurement uncertainties.

Verify the requirements of Specification 3.2.2.2 are satisfied.

c.

Satisfy the following relationship:

F (z) 5

) for P > 0.375 PxW(z) 1 l

F

N#)

F'(z) s O

_ for P 1 0.375 W(z) x 0.375 where F (z) is the measured Fn(z) increased by the agwances for manufact ring tolerances and me'&surement uncertainty, F is the F limit, K(z) is the normalized F (z) as a function of corh height, P ik n

the relative THERMAL POWER, and W(z) is the cycle-dependent function that accouats for pny' r distribution transients encountered during normal operation.

F

. K(z), and W(z) are specified in the COLR as perSpecification6.h.l.6.

MeasuringFf(z)accordingtothefollowingschedule:

d.

(1) Upon achieving equilibrium conditions after exceeding by 10% or more of RATED THERMAL POWER, the THERMAL POWER at which F (z) was last determined,

or g

During power escalation at the beginning of each cycle, the power level may be increased until a power level for extended operation _has been achieved and power distribution map obtained.-

Millstone - Unit 3

-3/4 2-13 Amendment No, g,60 0011

POWER DISTRIBUTION LIMITS SURVEILLANCE REOUIREMENTS (Continued)

(2) At least once per 31 Effective Full Power D'ays, whichever occurs

first, e.

With the maximum value of F(z)

K(z) over the core height (z) increasing since the previous determination of F (z), either of the following actions shall be taken:

(1)

F$(z) shall be increased by 2% over that specified in Specifica-tTon 4.2.2.2.2c, or (2)

F$(z) shall be measured at least once per 7 Effective Full Power Days until two successive maps indicate that the maximum value of F (z)

K(z) over the core height (z) is not increasing, f.

With the relationships specified in Specification 4.2.2.2.2c not being satisfied:

(1) Calculate the maximum percent over the core height (z) that F (z) exceeds its limit by the following expression:

q F (z) x W(z)

RTP

-1 x 100 for P 1 0.375 0

x K(z)

P i

Millstone - Unit 3 3/4 2-14' Amendment No. 77, )p-,60 oo11 j

I

w POWER DISTRIBUTION LIMITS SMRVEILLANCE RE0VIREMENTS (Continued) r f (2) x W(z)

-1 x 100 for P < 0.375 RTP 0

xK(z) 0.375 (2) One of the following actions shall be taken:

(a) Within 15 minutes, control the AFD to within new AFD limits which are determined by reducing the applicable AFD limits by 1% AFD for each percent F (z) exceeds its limits as o

determined in Specification 4.r.2,2.2f 1.

Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , reset the AFD alarm setpoints to thest modified limits, or (b) Comply with the requirements of Specification 3.2.2.2 for F (z) exceeding its limit by the percent calculated, or q

(c) Verify ti.at the requirements of Specification 4.2.2.2.3 for base load operation are satisfied and enter base load operation, i

g.

The limits specified in Specifications 4.2.2.2.2c, 4.2.2.2.2e, and 4.2.2.2.2f are not applicable in the.following core plane regions:

(1)

Lower core region from 0% tn 15%, inclusive.

(2) Upper core region from 85% to 100%, inclusive.

4.2.2.2.3 Base load operation i s permitted at powers above APL if the ND following conditions are satisfied:

Prig to entering base load operation, maintain THERMAL POWER above a.

APL and less than or equal to that allowed by Specifica-tion 4.2.2.2.2 for at least the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Maintain base load operation surveillance (AFD within the target band limit about the target flux difference of Specification -3.2.1.2)- during this time period.

Base load operation g then pgmitted providig THERMAL POWER is maintained between APL and APL or between APL and 100% (whichev-er is most limiting) and FnBgune ance is maintained pm tuant to Specification 4.2.2.2.4.

APC is defined as the minimum value ef:

l

(

l Millstone - Unit 3 3/4 2-15 Amendment No. #, 15,60 0011

,.-ce e

r*-.e h

P-

~e-e--ww--

-+c=-.--

,-=w...-

vg---

-,e+

t-r-,v-+-s

-e---

~vv+-++s

- - -,+-r-

4 POWER DISTRIBUTION LIMITS SURVEllLANCE RE0VIREMENTS (Continued)

OL APL 0

x 100%

N F (z) x H z)g(_

over the core height (z) where:

Ff(z)isthemeasuredF(z) increased n

by the allowances for manufac ing tolerances and measurement uncertainty.

The F limit is Fn W(Z)BL is the cycle dependent 0

function that acc.oonts for lirift ted powegp distribution transient encountered during base load operation.

specified in the COLR as per Specification J).9.1.6., K(z), and W(z)BL are b.

Durgg base load c,,eration, if the THERMAL POWER is decreased below APL then the conditions of 4.2.2.2.3.a shall be satisfied before reentering base load operation.

4.2.2.2.4 During base load operation F (z) shall be evaluated to determine if F (z) is within its limit by:

q n

Using the movable incore deteegrs to obtain a power distribution map a.

at any THERMAL POWER above APL b.

Increasing the measured F (z) component of the power distribution map n

by 3% to account for mant7facturing tolerances and further increasing the value by 5% to account for measurement uncertainties.

Verify the requirements of Sp.cification 3.2.2.2-are satisfied..

c.

Satisfying the following relationship:-

RTP F

l F(z)1 0 x K(z) for P > APLND P x W(z)BL M

RTP where:

F The F is the F limit, the normalizeh(z) is the measured F (z).F (z) as a funct P is khe relative n

THERMAL POWER.

W(z) is the cycle-dependent function that accounts for limited pgr diskribution transients encountered during base load operation.

F

, K(z), and W(z)BL are specified in the COLR as per Specification

.9.1.6.

Measuring F$(z) in conjunction with target flux difference determina-d.

tion according to the following schedule:

\\

l Millstone - Unit 3 3/4 2-16 Amendment No. J/, #.60 C011

.~

POWER DISTRIBUTION llMITS SURVElllANCE REQUIREMENTS (Continued)

(1)

Prior to entering base load operation after satisfying Sec-a tion 4.2.2.2.3, unless a full core flux map has been taken in the previous 31 Effective Full Power Days gth the relative THERMAL POWER having been maintained above APL for the 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to mapping, and (2) At least once per 31 Effective Full Power Days, e.

With the maximum value of F (z)

K(z) over the core height (z) increasing since the previous determination of F (z), either of the following actions shall be taken:

(1)

F (z) shall be increased by 2 percent over that specified in 4 2.2.2.4.c, or (2)

Ff(z) shall be measured at least once per 7 Effective Full Power Days until 2 successive maps indicate that the maximum value of F (z)

K(z) over the core height-(z) is not increasing, f.

With the relationship specified in 4.2.2.2.4.c not being satisfied, either of t'ne following actions shall be taken:

(1)

Place core in an equilibrium conditi where the limit in 4.2.2.2.2.c is satisfied, and remeasure F (z), or (2) Comply with the requirements of Specification 3.2.2.2 for F (z) 9 exceeding its limit by the maximum percent calculated over the core height (z) with the following expression:

Millstone - Unit 3 3/4 2-17 Amendment No. A7, #,60 0011

.... ~_.. -..

= -. -

POWER DISTRIBUTION LIMITS SURVEILLANCE REOUIREMENTS (Continued)

M F (z) x W(z)BL 0

ND

-1 x 100 for P 2 APL g

p xK(x) g.

The limits specified in 4.2.2.2.4.c, 4.2.2.2.4.e, and 4.2.2.4.f are not applicable in the following core plane _ regions:

(1)

Lower core region 0% to 15%, inclusive.

(2) Upper core region 85% to 100%,_ inclusive.

4.2.2.2.5 When F (z) is measured for reasons other than meeting the require-n ments of Specification 4.2.2.2.2, an overall measured Fn(z) shall be obtained from a power distribution map and increased by 3% to acclunt for manufacturing tolerances and further increased by 5% to account for measurement uncertainty.

(

1 l

l Millstone - Unit 3 3/4 2-18 Amendment No. U, g,60 0011

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

POWER DISTRIBUTION LIMITS 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR FOUR LOOPS OPERATING LIMITING CONDITION FOR OPERATION i

3.2.3.1 The indicated Reactor Coolant System (RCS) total flow rate and F shall be maintained as follows:

g a.

RCS total flow rate 2 387,480 gpm, and P [i.0 + PFAH (1.0 - P))

b.

F gsFAfi Where:

THERMAL POWER 1)

P

=

RATED THERMAL POWER 2)

FfH = Measured values of FfH obtained by using the movable incore detectors to ptain a power distribution map.

The measured value of F

should be used since Specifica-g tion 3.2.3.lb.takesinfoconsiderationameasurementuncertainty of 4% for incore measurement, RTP N

3)

F

- The F LkNITS REPORfH(limit at RATED THERMAL POWER COLR),

anhH - The power factor multiplier for FfHprovided in the COLR, l

4)

PF 5)

The measured value of RCS total flow rate shall be used since uncertainties of 2.4% for flow measurement have been included in Specification 3.2.3.la.

APPLICABILITY: MODE 1.

ACTION:

With the RCS total flow rate or F outside'the region of acceptable operation:

g a.

Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either:

1.

Restore the RCS -total flow rate and Ffg - to within the above limits, or l

2.

-Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux - High Trip Setpoint 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 .

MILLSTONE - UNIT 3 3/4 2-19 Amendment U,. JJ,60 0011

POWER DISTRIBUTI0ldlfiLLS LIMITING CONDITION FOR OPERATION ACTION (Continued) b.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of initially being outside the above limigs, verify through incore flux mapping and RCS total flow rate that F total flow rate are restored to within the above limits, AH and RCS or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

Identify and correct the cause of the out of-limit condition prior to c.

increasing THERMAL POWER above the reduced THERMAL POWER limit required by ACTION a.2, andpr b.,

above; subsequent POWER OPERATION may proceed provided that F and indicated RCS total flow rate are AH demonstrated, _ through incore flux mapping and RCS total flow rate comparison, to be within the region of acceptable operation prior to exceeding the following THERMAL POWER levels:

1.

A nominal 50% of RATED THERMAL POWER, 2.

A nominal 75% of RATED THERMAL POWER, and 3.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of attaining greater than or equal to 95% of RATED THERMAL POWER.

SURVEILLANCE REOUIREMENTS 4.2.3.1.1 The provisions of Specification 4.0.4 are not applicable.

4.2.3.1.2 RCS total flow rate and F shall bc determined to be within the H

acceptable range; Prior to operation above 75% of RATED THERMAL POWER after 9ach fuel a.

loading, and b.

At least once per 31 Effective Full Power Days.

4.2.3.1.3 The indicated RCS total flow rate shall be verified to be within the acceptable range at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the most recently obtained N

value of FAH, obtained per Specification 4.2.3.1.2, is assumed to exist.

4.2.3.1.4 The RCS total flow rate indicators shall be subjected to a CHANNEL CAllBRATION at least once per 18 months.

The measurement instrumentation shall be calibrated within 7 days prior to the performance of the calorimetric flow measurement.

MILLSTONE - UNIT 3 3/4 2-20 Amendment No. 60 0011

'l l

EQWER DISTRIBUTION LIMITS SURVEILLANCE REOUIREMENTS (Continued) 1 4.2.3.1.5 The RCS total flow rate shall be determined by precision heat balance C

measurement at least once per 18 months.

Within 7 days prior to performing the precision heat balance, the instrumentation used for determination of steam pres,ure, faadwater pressure, feedwater temperature, and feedwater venturi AP in the calorimetric calculations shall be calibrated.

4.2.3.1.6 If the feedwater venturis are not inspected at least once per 18 months, an additional 0.1% will be added to the total RCS flow measurement uncertainty, f

l l

MILLSTONE - UNIT 3 3/4 2-21 Amendment No. 27,60 ooit l...._.....,___

,__...,_,._m_,__

m..

,,c

,.,,.r,

~l EQ Q DISTRIBUTION LIMITS RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR THREE LOOPS OPERATING LIMITING CONDITION FOR QPERATION t

The indicated Reactor Coolant System (RCS) total flow rate and Ffg 3.2.3.2 shall be maintained as follows:

a.

RCS total flow rate 2 303,200 gpm, and N

b.

F IFRTP (1.0 + PFAH (1.0 - P)]_

3H 3

Where:

1)

THERMAL PCWER P

=

RATED THERMAL POWER N

N 2)

F Measured values of F obtained by using the movable ibUo=redetectorstoobtainNpowerdistributionmap.

N The measured value. of F should be used since Speci-fication3.2.3.2b.takesintYconsiderationameasure-ment uncertainty of 4% for incore measurement, RTP 3)

F The F limit e+ RATED THERMAL POWER in t' e CORE

=

h ONRATINGLIMINREPORT t),

4)

PFAH = The power factor. J tplier for F in the COLR, and gg S)

The measured value of RCS xal flow rate shall be used since uncertainties of 2.8% for flow measurement have been included in Specification 3.2.3.2a.

APPLICABILITY: MODE 1.

ACTION:

With the RCS total flow rate or F outside the region of acceptable operation:

AH a.

Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either:

l.

Restore the RCS total flow rate and F to within the above limits, or g

2.

Reduce THERMAL POWER to less than 32% of RATED THERMAL POWER-and reduce the Power Range Neutron Flux - High Trip Setpoint to less than or equal to 37% of R/JID THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

MILLSTONE - UNIT 3 3/4 2-22 Amendment No. #, 59,60 0011

]

POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPEPATION ACTION (Continued) b.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of initially being outside the above limits, verify through incore flux mapping and RCS total flow rate that FfH and RCS total flow rate are restored to within the above limits, or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , Identify and correct the cause of the out-of-limit condition prior c.

to increasing THERMAL POWER above the reduced THERMAL POWER limit required by ACTION a.2. and/or b.,

above; subsequent POWER OPERATION may proceed provided that F and indicated RCS total flow raw are AH demonstrated, through incore flux mapping and RCS total flow rate comparison, to be within the region of acceptable operation prior to exceeding the following THERMAL POWER levels:

1.

A nominal 22% of RATED THERMAL POWER, and 2.

A nominal 50% of RATED THERMAL POWER.

SURVEILLANCE RE0VIREMENTS 4.2.3.2.1 The provisions of Specification 4.0.4 are not applicable.

4.2.3.2.2 RCS total flow rate and F shall be determined to be within the AH acceptable range at least once per 31 Effective Full Power Days.

4.2.3.2.3 The indicated RCS total flow rate shall be verified to be within the acceptable rangg at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the most recently obtained value of F exist.

AH, btained per Specification 4.2.3.2.2, is assumed to 4.2.3.2.4 The RCS total flow rate indicators shall be subjected to a CHANNEL CAllBRATION at least once per 18 months.

The : measurement instrumentation shall be calibrated within 7 days prior to the performance of the calorimetric flow measurement.

4.2.3.2.5 The RCS total flow rate shall be determined by precision heat balance measurement at least once per 18 months.

Within 7 days prior to pe-forming the precision heat

balance, the instrumentation used for-determination of steam pressure, feedwater pressure, feedwater temperature, and feedwater venturi AP in the calorimetric calculations shall be calibrated.

4.2.3.2.6 If the feedwater venturis are not inspected at least once per 18 months, an additional 0.1% will be added to the total RCS flow measurement uncertainty.

MILLSTONE - UNIT 3 3/4 2-23 Amendment No. //,60 0011

i POWER DISTRIBUTION LIMITS 3/4.2.4 OUADRANT POWER' TILT RATIO -

LIMITING CONDITION FOR-0PERATION 3.2.4 The QUADRANT POWER TILT RATIO shall not exceed 1.02.-

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

4 ACTION:

a.

With the QUADRANT - POWER ' TILT RATIO determined -to exceed - 1.02 but :

less than or equal to-1.09:

1.

Calculate the QUADRANT POWER TILT RATIO.at least once per hour until-either:

a), The QUADRANT -POWER TILT RATIO --is reduced to within -~its-limit, or b)

THERMAL POWER is reduced to less than 50% of RATED THERMAL -

POWER.

2.

Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months _ei_ther:

s a)

Reduce the QUADRANT POWER' TILT RATIO to_within its limit, or b)

Reduce THERMAL ~ POWER at least 3% from RATED ' THERMAL < POWER-for - each E l% of ' indicated -QUADRANT _ POWER. TILT-RATIO in excess of 1 and-similarly reduce the Power Range Neutron i

Flux-High Trip Setpoints within the next 4-hours, 3.

Verify that the QUADRANT:- POWER-TILT RATIO is: within-its limit-within '24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after. exceeding the~ limit or. reduce THERMAL POWER to less than 50% of RATED-THERMAL-POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and ' reduce : the-Power Range '. Neutron' Flux-High Trip 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 4.

Identify andDcorrect the cause of the out-of-limit condition prior to increasing THERMAL. POWER; subsequent POWER OPERATION above 50% of RATED THERMAL: POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within:its limit at least-once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or. until verified acceptable at 95% -

or greater RATED THERMAL POWER.

See Special Test Exceptions Specification 3.10.2.

MILLSTONE - UNIT 3 3/42-24L Amendment No. 60-0011

-i POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION ACTION (Continuedl b.

With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to misalignment of either a shutdowis or control rod:

1.

Calculate the QUADRANT POWER TILT RATIO at least once per hour until either:

a)

The QUADRANT POWER TILT RATIO is reduced to within its limit, or b)

THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER.

2.

Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1, within 30 minutes; 3.

Verify that the QUADRANT POWER TILT RATIO is within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after exceeding the limit or reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High Trip 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 4.

Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified acceptable at 95%

or greater RATED THERMAL POWER.

With the QUADRANT POWER TILT RATIO determined to exceed 1,09 due to c.

causes other than the misalignment of either a shutdown or control rod:

1.

Calculate the QUADRANT POWER TILT RATIO at least once per hour until either:

a)

The QUADRANT POWER TILT RATIO is reduced to within its limit, or b)

THERFML POWER is reduced to less than 50% of RATED THERMAL POWER.

MILLSTONE - UNIT 3 3/4 2-25 Amendment No. 60 0011

. _. _... _. _ _ _.~

i_

i POWER DISTRIBUTION LIMIJ1-LIMITING CONDITION FOR OPERATION ACTION (Continued) 2.

Reduce-THERMAL POWER to less than 50% of RATED THERMAL POWER within : 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ande reduce the Power Range Neutron Flux High Trip 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.

Identify and correct -the cause of the-out-of limit condition-prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least t

once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until-- ver_ified at 95% or greater-RATED THERMAL-POWER.-

SURVEILLANCE REOUIREMENTS 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 OPEr ABt.E, and b.

Calculating the ratio at least once per 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s: during steady-state operation when-the alarm is -inoperable.-

4.2.4.2 The QUADRAN1 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 confirm _that the normalized -symmetric power distribution,- obtained fror two sets of. four_-

symmetric thimble locations.or-full-core flux map, is we,istent with the indicated QUADRANT POWER TILT RATIO at least~once per-12 hours.

MILLSTONE - UNIT'3 3/4 2-26 Amendment No. p,60 0011

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

n Aff; 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 limits shown on Table 3.2-1:

Reactor Coolant System T,yg, and a.

b.

Pressurizer Pressure.

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 .

SURVEILLANCE REQUIREMENTS 4.2.5 Each of the parameters of Table 3.2-1 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 .

MILLSTONE - UNIT 3 3/4 2-27 Amendment No. p,60 0011

_________________--------------I

TABLE 3.2-1 33 DNB PARAMETERS

%F S

E LIMITS m

Three Loops iri Opera-E Four Loops in tion & Loop Stop p

PARAMETER Operation Valves Closed w

Indicated Reactor Coolant System T,yg 5 591.l*F 1 583.3*F Indicated Pressurizer Pressure 2 2218 psia

1 2218 psia

l

~R.

7 El l

)

E 2

. i.'

F N

'm

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.

A

~ ' ~

i TABLE 3.3-1 i

i5

)

REACTOR TRIP SYSTEM INSTRUMENTATION o r-

"C

-4 o

MINIMUM E

TOTAL NO.

CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS JO TRIP OPERABLE MODES ACTION C5 1.

Manual Reactor Trip 2

1 2

1, 2 1

-4 2

1 2

3*,

4*, 5*

11 l

w 2.

Power Range, Neutron Flux A

a.

High Setpoint 4

2 3

1, 2 2

b.

Low Setpoint 4

2 3

l###, 2 2

3.

Power Range, Neutron Flux 4

2 3

1, 2 2

High Positive Rate 4.

Power Range, Neutron Flux, 4

2 3

1, 2 2

High Negative Rate w

=

w 5.

-Intermediate Range, Neutron Flux 2

1 2

l###, 2 3

4 6.

-Source Range, Neutron Flux a.

Startup 2

1 2

2##

4 b.

Shutdown-2 1

2 3*,

4*, 5*

11 7.

Overtemperature AT a.

Four Loop Operation-4 2

3 1, 2 6

F b.

Three Loop Operation 3

2 2

1, 2 6

i

~~1868.5 psig~~

~~2) Channel III and IV 22.16 15.6 3.3 2 1877.3 psig 2 1863.3 psig e.~~
Steam Line Pressure--Low 17.7 15.6 2.2 2 658.6 psig*
2 648.3 psig*
2.
Containment Spray (CDA) i a.
Manual Initiation N.A.
N.A.
N.A.
N.A.
N.A.
b.
Automatic Actuation' Logic N.A.
N.A.
N.A.
N.A.
N.A.
i and Actuation Relays i
![
c.
Contair. ment Pressure--High-3 3.3 1.01 1.75 s 8.0 psig s 8.8 psig j
Rg 3.
Containment Isolation a.
Phase "A" Isolation z
P i
~~1) Manual Initiation N.A.~~
N.A.
N.A.
N.A.
N.A.
)
i L
n-REACTOR COOLANT SYSTEM b
i HOT STANDBY llMITING CONDITION FOR OPERATION p
3.4.1.2 At least three of the reactor coolant loops listed below shall be OPERABLE, with at least three reactor coolant loops in operation when the Reactor Trip System breakers are closed or with at least one reactor coolant loop in operation when the Reactor Trip System breakers are open:*
a.
Reactor Coolant Loop 1 and its associated steam generator and reactor coolant pump, b.
Reactor Coolant loop 2 and its associated steam generator und
+
reactor coolant pump, c.
Reactor Coolant Loop 3 and its associated steam generator and reactor coolant pump, and d.
Reactor Coolant Loop 4 and its associated steam generator and reactor coolant pump.
APPLICABILITY: H0DE 3.
ACTION:
With less than the above required reactor coolant loops OPERABLE, a.
restore the required loops to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
b.
With less than the above required reactor coolant loops in operation and the Reactor Trip System breakers in the closed position, within I hour open the Reactor Trip System breakers.
With no reactor coolant loop in operation, suspend all operations c.
involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required reactor coolant loop to operation.
SURVElllANCE RE0VIREMENTS 4.4.1.2.1 At least the above required reactor coolant aucps, if not in operation, shall be determined OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability.
4.4.1.2.2 secondary side water level to be greater 'than or equal-to 17% at leas per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
4.4.1.2.3 The required reactor-coolant loops shall be verified in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
~~All reactor coolant pumps may be deenergized for up to I hour provided:~~
(1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 100F below saturation temperature.
MILLSTONE - UNIT 3 3/4 4-2 Amend,nent-No. 60 0019
l REACTOR COOLANT SYjil8 l
COLD SHUTD,2WN - LOOPS NOT FILLED
,llMITING CONDITION FOR OPERATION
~'
3.4.1.4.2 Two residual heat removal (RHR) loops shall be DPERABLE* and at l
least one RHR loop shall be in operation.** The chenical and volume control system (CVCS) shall be aligned to preclude Reactor Coolant System ooron concentration reduction or the SHUTDOWN MARGIN of Specification 3.1.1.2 shall be met, i
APPLICADil11Y:
MODE 5 with less than two reactor coolant loops filled, l
ACTION:
With less than the above required RHR loops OPERABLE, imediately a.
initiate corrective action to return the required RHR loops to OPERABLE status as soon as possible.
b.
With no RHR loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and I
immediately initiate corrective action to return the required RHR loop to operation, With the CVCS dilution flow paths not closed and secured in c.
position, imediately close and secure the paths or satisfy the SHUTDOWW MARGIN of Specification 3.1.1.2.
EURVEILLANCE R @jlREMENTS 4.4.1 4.2.1 The required RHR loops shall be demonstrated OPERABLE pursuant to Specification 4.0.5.
4.4.1.4.2.2 At least one RHR loop shall be determined to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
4.4.1.4.2.3 At least per 31 days the following valves shall be verified closed and locked.
The valves may be opened on an intermittent basis under administrative control.
0ne RHR loop may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing provided the other RHR loop is OPERABLE and in operation.
~~The RHR pump may be deenergized for up to I hour provided:~~
(1)noopera-tions are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 100F below saturation temperature.
MILLSTONE - UNIT 3 3/4 4-6 0036 Amendment No. 60
i
.y SURVE{LLANCE REQUIREMESTS (Con't) yl ve Number Valve Function Valve Positio/1 l
1.
V304(I-)
Primary Grade Water Closed to CVCS 2.
V120(Z-)
Mderating Hx Outlet Closed 3.
V147(Z-)
itRS 0;ttlet C10 sed 4.
V797(Z-)
Failed Fuel Monitoring Closed Flushin S.
FIS0(Z-)
Resin S ulce, CVCS Cation Closed l
Bed Demineralizer 6.
t'571(2-)
Rosin S M ci, CVCS Cation Cit M Bed Deminurilizer 7,
V111(Z )
Rnin Sluica, CVCS Cation Closed bed Demineralizer 8.
Vll2(Z-)
Resin S'iuice, CVCS Cation Closed Bed Demineralizer 9.
V98(Z-)/V99(Z-)
Resin Sluice, CVCS Mixed Closed Bed Demlieralizer
~~10. VS69(Z-)/V570(Z-)~~
Resin Sluicu, CVCS Mixec' Closed
)
1 E2d Demineraliter 11.
V107(Z-)/V109(Z-)
Resin Sluice, CVCS Mixed Cicsed Bed Deafneralizer
~~12. V108(2-)/V110iZ-)~~
Resin Sluice, CVCS Mixcd Closed l
Bed Demineralizer i
l l
l MILLSTONE - UNIT 3 3/4 4-6a Amendment No. 60 2036
.\\
- - - ~ ~ ~ ~
PEACTOR COOLANT SYSTEM ISOLATED LOOP STARTMD, LINITING CONDITION FOR OPERAT107,_
)
3.4.1.6 A reactor coolant irr,9 rW1 remain istbted with power removed from the assoc'rted PCS 'icca stg wive operators until:
4 a.
The t.emperature at the cold 1cp 4f the isolated loop. is within 20*I of tha highsst cold le7 ti.operature V the operating loops, t.
The non concuiration ef the isolated bop is greater than or equal to st/c boron concentration r,' tne operating loops, or greater thin 2600 ppm wiichever is lens, y
me isolated porner of tp loop hu been th.Ined aW M Milled, c
and d.
The reactor is subcritical by at least te value
'e mo4 by Specificationi 3 1.1.1.2 or 3.1.1.2 for Mode b or SpdC','~ cation 3.9.1.1 for Mode 6 APPLICABILITY: MODES 5 and 6.
Ell @
With the requirements of the above specification not satisfied, do a.
not open the isolated loop stop valves.
SURVEILLANCE REOUIREMENTS
4. 4.1. 6.1 The isolated loop cold leg temperature shall be deterr(ned te se within 20'F of the highest cold leg ter jerature of the operating loops within 30 minutes prior to opening the cold ic.g stop valve.
4.4.1.6.2 The reactor shall be determined to be subcritical by Eat least the value required by Specifications 3.1.1.1.2 or 3.1.1.2 for Mode 5 or Specification 3.9.1,1 for Mode 6 within 30 minutes prior to. opening the cold leg stop tnive.
4.4.1.6.3 Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to opening the loop stop valves, the isolated loop shall be determined to:
a.
Bo drained and refilled, and b.
Have a boron concentration greater than or equal to the baron concentration of the operating loops, or greater than 2600 ppm -
whichever is less.
HRI. STONE - UNIT 3 3/4 4-8 Amendment No. JJ, 57,60 0020
1 E
O g
03 250 i
.)
D 5l' I
UNACCEPTABLE W
200 OPERATION l
u 8
Fo 150 E
a 8
u e
y 100 E
a.
9 ACCEPTABLE 50 OPERATION 5
t g
w N
O
~
8 20 30 40 50 60 70 80 90 100 i
PERCENT OF RATED THERMAL POWER FIGURE 3,4-1 DDSE EQUIiALENT l-331 REACTOR COOLANT SPECIFIC ACTIVITY LIM:T VERSUS PERCENT Of RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVlY >l um/ gram DOSE EQUIVALENT l-131 l
I WILLST0E - UNIT 3 3/4 4-30 Amendment No. 60 l
- - - -- ~ ~
1 MATERIAL PROPERTY, BASIS CONTROLLING MATERIAL
~~PLATE METAL 4~~
~~COPPER CONTENT~~
~~CONSERVATIVELY ASSUMED TO BE 0.10 WT %~~
~~PHOSPHORUS CONTENT~~
~~0.010 WT %~~
~~RTNDT INITIAL~~
~~GOT RTNDT AFTER 10 EFPY~~
1/4T.122*F 3/4T 101*F CURVE APPLICABLE FOR HEATUP RATES UP TO 607/HR FOR TE SERVICE PERIOD UP TO 10 EFPY AND CONTAINS MARGINS OF 10*F AND 60 PSIG FOR POSSIBLE INSTRUMENT ERRORS 3000.0 LEAK TEST N LIMIT a
/
/
g 2000,0 w
0 E
l
<r i
S 1000.0 HEATUP CURVE
/
CRITICALITY LIMIT
' BASED ON INSERVICE HYDROSTATIC TEST TEMPERATURE (266 *F )
FOR THE SERVICE PERIOD UP TO 10 EFPY 0.0 0.0 100.0 200.0 300.0 400.0 500.0 INDICATES TEMPERATURE (DEG. F)
FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS - APPLICABLE UP TO 10 E7PY WILLST0E - UNIT 3 3/4 4-34 Amendment No '60
,, 1 MATERIAL PROPERTY BASIS CONTROLLING MATERIAL
~~DLATE METAL COPPER CONTENT~~
~~CONSERVATIVELY ASSUMED 10 BE 0.10 WT %~~
~~PHOSPHORUS CONTENT~~
~~0.010 WT %~~
~~RTNOT INITIAL~~
~~607 RTN D',~~
~~AFTER 10 EFPY~~
~~1/4T 1227 3/4T,1017 l~~
CURVE APPLICABLE FOR COOLDOWN RATES UP TO 1007/HR FOR THE SERVICE PERIOD UP TO 10 EFPY AND CONTAINS MARGINS OF 10*F AND 60 PSIO FOR POSSIBLE lhSTRUMENT ERRORS 3000.0 I
2000.0 O
E E
e 1000.0 COOLDOWN RATES ('F/HR) 20 --w 100 0.0 0.0 100.0 200.0 300.0 400.0 500.0 t
l INDICATES TEMPERATURE (DEG. F)
FIGURE 3.4-3 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS - APPLICABLF, UP TO'10 EFPY ao
~
=-
P 800 3M
$b w
i5 2
Q 700 k
s W
8 3
~~d 600 il2W 5~~
$z 500 i
D2 E
1 50 100 200 300 400 TR1D - AUCTONEERED LOW WEASURED RCS TEMPERATURE (*F) l FIGURE 3.4-4o NOMINAL MAXIMUW ALLOWABLE PORV SETPOINT FOR THE COLD OVERPRESSURE SYSTEM (FOUR LOOP DPERATION)-
WILLSTOE - UNIT 3
-3/4 4-40 Amendment No. 60
- =.
l l
l 800 G
Vi&
6-E O
Ow 700 t-E s
l h
600 s
i 5
d I
bz 500 0
52 E
I 50 100 200 300 400 TRTD - AUCTONEERED LOW MEASURED RCS TEMPERATURE (T)
FICARE 3.4-4b NCMINAL MAXIMUM ALLOWABLE PORV CETPOINT FOR THE COLD OVERPRESSURE SYSTEM (THREE LOOP OPERATION) 1 WILLST0E - UNIT 3 3/4 4-41 Amendment No. 60
~
3/4.5 EMERGENCY CORE COOLING SYSTEMS i
2/J.5.1 AttuMULATOR,1 LIMITING CONDITION FOR OPERATION
}
3.5.1 Each Reactor Cuolant System (RCS) accumulator shall be OPERABLE with:
.he isolation valve open and pow'er removed, a.
b.
A contained borated water volume of between 6618 and 7030 gallons, A boron concentration of between 2600 and 2900 ppm, and c.
f d.
A nitrogen cover pressure of between 636 and 694 psia.
APPLICABILITY: MODES 1, 2, and 3*.
ACTION:
With one accumulator inoperable, except as a result of a closed a.
l isolation valve, restore the inoperable accumulator to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in at least HOT STAND 8Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce pressurizer pressure to less than 1000 psig within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , b.
With one accumulator inoperable due to the isolation valve being closed, either immediately open the italation valve or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce pressurizer pressure to less than 1000 psig within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
SURVEILLANCE REQUIREMENTS 4.5.1.1 Each accumulator shall be demonstrated OPERABLE:
a.
At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by:
1)
Verifying the contained borated water veiume and nitrogen cover pressure in the tanks to be within the above limits, and t
2)
Verifying that each accumulator isolation valve is open, i
b.-
At least once per 31 days and within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after each solution volume increase of greater than or equal to 1% of tank volume by verifying the boron concentration of the accumulator solution; and i
~~Pressurizer pressure above 1000 psig.~~
MILLSTONE - UNIT 3 3/4.5-1 Amendment No. JJ, JJ,60.
0021 l. - - -. - - - - - -
' - - - - - * ~ ~ ~ ~ ~ ~ ~ ' ~ ~ ~ ~ ~ ~ ' ~ ~ ~ ~ ~ ~ ~
EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REOUIREMENTS 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE:
At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that the following valves a.
are in the indicated positions with power to the valve operators removed:
i Valve Numb.gC Valve Function Valve Position 3SlH+MV8806 RWST Supply to S1 Pumps OPEN 3SlH*MV8802A S1 Pump A to Hot leg Injection CLOSED 3S!H*MV88028 S1 Fump D to Hot Leg injection CLOSED 3SlH*MV8835 SI Cold Leg Master isolation OPEN 3SlH*MV8813 S1 Pump Master Miniflow OPEN lsolation 3SIL*MVB840 RHR to Hot Leg injection CLOSED 3SIL*MV8809A RHR Pump A to Cold leg OPEN Injection 3SIL*MV8809B RHR Pump B to Cold leg OPEN i
Injection b.
At least once per 31 days by:
1)
Verifying that the ECc5 piping, except for the RSS pump, heat exchanger and associated piping, is full of water by venting the ECCS pump casings and accessible discharge piping high points, and 2)
Verifying that each valve (manual, power operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
1 c.
By a visual inspection which verifies that no loose debris (rags, trash, clothing, etc.) is present in the containment which could be transported to the containment sump and cause restriction of the aump suctions during LOCA conditions.
This visual inspection shall se performed:
1)
For all accessible areas of the containment prior to establish-ing CONTAINMENT INTEGRITY, and 2)
Of the areas affected within containment at the completion of each containment entry when CONTAINMENT INTEGRITY is established.
d.
At least once per 18 months by:
1)
Verifying automatic interlock action of the RHR System from the Reactor Coolant System by ensuring that with a simulated or actual Reactor Coolant System pressure signal' greater than or equal to 390 psia the interlocks prevent the valves from being opened.
l MILLSTONE - UNIT 3 3/4 5-4 Amendment No. 60
EMERGENCY CORE C00llNG SYSTEMS SURVElllANCE RE0VIREMENTS (Continued) 2)
A visual inspection of the containment sump and verifying thtt the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screehs, etc.
evidence of structural distress or abnormal corrosion.) show no At least once per 18 months, during shutdown, by:
e.
1)
Verifying that each automatic valve in the flow path actuates to its correct position on a Safety Injection-actuation test signal, s.nd 2)
Verifying that eac!: of the following pumps start automatically upon receipt of a Safety injection actuation test signal:
a)
Centrifugal charging pump, 1
b)
Safety injection pump, and c)
RHR pump.
3)
Verifying that the Residual Heat Removal pumps stop automatically upon receipt of a low Low RWST Level test signal, f.
By verifying that each of the following pumps develops the indicated differential pressure on recirculation flow when tested pursuant to Specification 4.0.5:
1)
Centrifugal charging pump-1 2411 psid, 2)
Safety injection pump 1 1348 psid, 3)
RHR pump 2 165 psid, and 4)
Containment recirculation pump 1 130 psid, g.
By verifying the correct position of each electrical and/or mechanical position stop for the following ECCS throttle valves:
1)
Within-4 hours 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.
l ECCS Throttle Va].yn Valve Number
. Valve Number
-- --3SlH*V6-
-3S!H*V25-3SlH*V7 3SlH*V27-MILLSTONE'- UNIT 3 3/4 5 5 Amendment No. 60 ooaa
-l j
E 1
{MERGENCY CORE COOLING SYSTEjg SURVEILLANCE RE0UIREMENTS (Continued) i ECCS Throttle Valves Valve Number Valve Numher 3SlH*V8 3S!H*V107 3SlH*V9 3S!H*V108 351H*V21 3S!H*V109 3SIH*V23 3S!H*V111 h.
By performing a flow balance test, during shutdown, following com-pletion of modifications to the ECCS subsystems that alter the subsystem flow chcraeteristics and verifying that:
1)
For centrifugal charging 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 339 gpe, and b)
The total pump flow rate is less than or equal to 560 gpm.
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 442.5 gpm, and b)
The-total pump flow rate is less than or e gpm for the A pump and 650 gpm for the B pump. qual to 670
-3)
For RHR pump lines, with a single pump running, the sum of the injection line flow rates is greater than or equal to 3976 gpm.
I i
4
)
MILLSTONE UNIT 3 3/456 Amendment No. 60 0033
. ~ -. -., -. _,. - - _ -, _.. -..... - -.. -.. - -...
i EMERGENCY CORE COOLING SYSTEMS 3/4.5.4 REFUELING WATER 510 RAGE TANK LlfilTING CONDITION FOR OPERATION 3.5.4 The refueling water storage tank (RWST) shall be OPERABLE with:
a.
A contained borated water volume between 1,166,000 and 1,207.000
~~gallons, b.~~
A boron concentration between 2700 and 2900 ppm of boron, A minimum solution temperature of 40'F, snd c.
d.
A maximum solution temperature of 50'F.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION With the RWST inoperable, restore the tank to OPERABLE status within I hour or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within t following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
SURVEILLANCE REOUIREMENTS 4.5.4 The RWST shall be demonstrated OPERABLE:
a.
At least once per 7 days by:
1)
Verifying the contained bor: ted water volume in the tank, and 2)
Verifying the boron concentration of the water, b.
At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature.
1 l
l MILLSTONE.- UNIT 3 3/4 5 9 Amendment No, JJ,60 00t3
1 4
i CONTAINMENT SYSTEMS SPRAY ADDITIVE SYSTEM LIMITING CONDITION FOR_ OPERATION i
3.6.2.3 The Spray Additive System shall be OPERABLE with:
A chemical addition tank containing a volume of between 17,760 and a.
18,760 gallons of between 3.4 and 4.1% by weight NaOH solution, and b.
Two gravity feed paths each capable of adding Na0H solution from the chemical addition tank to each Containment Quench Spray subsystem pump suction.
APPLICABILITY: MODES 1, 2, 3, and 4.
M11M:
With the Spray Mditive System inoperable, restore the system to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months st restore the Spray Additive System to OPERABLE status within the next 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months 4
or be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
S R yEILLANCE RE001REMENTS 4.6.2.3 The Spray Addithe System shall be demonstrated OPERABLE:
At least once per 31 days by verifying that each valve (manual, a.
power operated, or automatic) in the flow path that is not locked,
~~sealed, or otherwise secured in positior, is in its correct position; b.~~
At least once per 6 months by:
1)
Verifying the contained solution volume in the tank, and 2)
Verifying the concentration of the NaOH solution by chemical analysis i.t witMn the above limits.
At least once per 18 months, during shutdown, by verifying that each c.
automatic valve in the flow path actuates to its correct position on a CDA test signal.
l l
L l
l MILLSTONE - UNIT 3 3/4 6 14.
Amendment No.- #.60-0024
. - - _ - ~ ~ -, _. - -. - _ -. _ _ ---_.-_ -.,.
'f 3/4.9 REFUELING OPERAT10h5 l
i 3/4.9.1 BORON CONCENTRATION
{
L[dHjnG CONDITION FOR OPERATION l
1 3.9.1.1 The boron concentration of all filled portions of the Reactor Coolant t
System and the refueling canal shall be maintained uniform and sufficient to i
ensure that the more restrictive of the following reactivity conditions is i
met; either:
i A K,gg of 0.95 or less, or a.
b.
A boron concentration of greater than or equal to 2600 ppm.
i Additionally, the CVCS valves of Specification 4.4.1.4.2.3 shall be closed and secured in position.
APPLICABillTY: MODE 6.*
ACTION:
With the requirements of the above specification not a.
satisfied, immediately suspend all operations involving CORE ALTERATIONS or i
positive reactivity changes and initiate and continue boration at greater than or equal to 33 gpm of a solution containing greater i
than or equal to 6300 ppm boron or its equivalent until K is reduced to less than or equal to 0.95 or the boron concentrat k is i
restored to greater than or equal to 2600 ppm, whichever is the more ratrictive.
i b.
With any of the CVCS valves of Specification 4.4.1.4.2.3 not I
closed ** and secured in position, immediately close and secure the valves.
SURVEILLANCE RE0VIREMENTS 4.9.1.1.1 The more restrictive of the above two reactivity conditions shall be determined prior to:
Removing or unbolting the reactor vessel head, and a.
i b.
Withdrawal of any full-length control rod in excess of 3 feet from j
its fully inserted position within the reactor vessel.
~
4.9.1.1.2 The boron concentration of the Reactor Coolant System and the refueling canal shall be determined by chemical analysis at -least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , t
4.9.1.1.3 The CVCS valves of Specification 4.4.1.4.2.3 shall be verified closed and locked at least once per 31 days.
}
~~The reactor shall be maintained in MODE 6 whenever fuel is in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.~~
~~Except-those opened under administrative control.~~
MILLSTONE - UNIT 3 3/4 9 1 Amendment No. g,60 0025 I
_.._,_,,,,,,,-...._.m
- - ~.
i i
i 3/4.1 PEACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 B0 RATION CONTROL f
i 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN the re) activity transients asso-the reacf A sufficient SHUTDOWN MARGIN ensures _that:
(1 suberitical from all operating conditions, 2
ciated with postulated accident conditions (ar)e controllable within acceptable-limits, and (3 preclude inadve)rtent criticality in the shutdown condition.the reactor will be ma i
SHUTDOWN MARGIN requirements vary throughout core life as, function of fuel depletion, RCS boron concentration, and RCS T In MODES 1 knd 2, the most restrictive condition occurs at E0L with fV9, at no load operating temperature, and is associated with a resulting uncontrolled RCS cooldown, postulated stN8 line break accident and i
in the analysis of this accident, a i
minimum SHUTDOWN MARGIN of.l.3% AK/K is required to control the reactivity l
transient.
Accordingly, the SHUTDOWN MARGIN requirement is based upon this limiting condition and is consistent with FSAR. safety analysis assumptions.
in MODES 3, 4 and 5, the most restrictive condition occurs at BOL, associated with a boron dilution accident.
In the -analysis of this accident, a~ minimum SHUTDOWN MARGIN as defined in Specification 3/4.1.1.2 is required to allow the operator 15 minutes from the initiation of the Shutdown Margin Monitor alarm i
i to total loss of SHUTDOWN MARGIN.
Accordingly, the SHUTDOWN MARGIN requirement is based upon this limiting requirement and is consistent with the i
accident analysis assumptions.
The required SHUiDOWN MARGIN is plotted as a j
function of RCS critical boron concentration, t
3/4.1.1.3 MODERATOR TEMPERATURE COEFFICIEtil The limitations on moderator temperature coefficient- (MTC) are provided to ensure that the value of this coefficient remains within the limiting
[
condition assumed in the FSAR accident and transient analyses.
The MTC values of this specification are applicable to a specific set of plant conditions; accordingly, verification of MTC values at. conditions other than those explicitly stated will require extrapolation to those conditions in i
L order to pormit.an accurate comparison.
L The most negative MTC, value equivalent to the most positive moderator density coefficient (MDC), was obtained by incrementally correcting the MDC used in the FSAR analyses to nominal operating conditions.
i 1
MILLSTONE - UNIT 3 B 3/4 1-1 Amendment No. J,60
-0026
_,... - _ _... _ _ _ _ _ _ ~, _
(
REACTIVITY CONTROL SYSTEMS BASES 4
BORATIONSYSTEMS(Continued)
MARGIN from expected operating conditions of equivalent to that required by Figure 3.1-5 after xenon decay and cooldown to 200*F.
boration capability requirement occurs at EOL from fullThe maximum expected aower equilibrium xenon conditions and requires a usable volume of 21,020 gallons of 6300 ppm borated water from the boric acid storage tanks or 1,166,000 gallons of 2700 ppm borated water from the refueling water storage tank RWST).
A gallons is specified to be consi(stent with minimum RWST volume of 1,166,000 ECCS requirement.
With the RCS temperature below 200'F, one Baron Injection System is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single Boron injection System becomes inoperable.
The limitation for a maximum of one centrifugal charg ng pump to be OPER-ABLE and the Surveillance Requirement to verify all charg ng pumps except the required OPERABLE pump to be inoperable below 350'F provi es assurance that a mass addition pressure transient can be relieved by the operation of a single PORV.
The boron capability required below 200'F is sufficient to provide a SHUTDOWN MARGIN of 1.3% Ak/k after xenon decay and cooldown from 200'F to 1400F, This condition requires either a usable volume of 4100 gallons of 6300 ppm borated water from the boric acid storage tanks or 250,000 gallons of 2700 ppm borated water from the RWST.
The unusable volume in each boric acid storage tank is 1300 gallons.
The contained water volume limits include allowance for water not available because of discharge line location and other physical characteristics.
The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between 7.0 and 7.5 for the solution recirculated within containment after a LOCA.
This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.
The minimum RWST solution temperature for MODES 5 and 6 is based on analysis assumptions in addition to freeze protection considerations.
The minimum / maximum RWST solution temperatures for MODES 1, 2, 3 and 4 are based on analysis assumptions.
The OPERABILITY of one Boron injection System during REFUEllNG ensures that this system is available for reactivity control while in MODE 6.
MILLSTONE - UNIT 3 B 3/4 1-3 Amendment No. L2,60 00t7
REACTIVITY CONTROL SYSTEMS BASES BQYABLECONTROLASSEMBLIES(Continued) rod alignment and insertion limits.
Verification that the Digital Rod Position Indicator agrees with the demanded position within 112 steps at 24, 48,120, and fully withdrawn position for the Control Banks and 18, 210, and fully withdrawn position for the Shutdown Banks provides assurances that the Digital Rod Position Indicator is operating correctly over the full range of i
indication.
Since the Digital Rod Position Indication System does not indicate the actual shutdown rod position between 18 steps and 210 steps, only points in the indicated ranges are picked for verification of agreement with demanded position.
The ACTION statements which permit limited variations from the basic requirements are accompanied by additional restrictions which ensure that the original design criteria are met.
Misalignment of a rod requires measurement of peaking factors and a restriction in THERMAL POWER.
These restrictions provide assurance of fuel rod integrity during continued operation.
In addition, those safety analyses affected by a misaligned rod are reevaluated to confirm that the results remain valid during future operation.
The maximum rod drop time restriction is consistent with the assumed rod drop time used in the safety analyses.
Measurement with T greater than or equal to 5510F and with all reactor coolant pumps operatidd9 ensures that the measured drop times will be representative of insertion times experienced during a Reactor trip at operating conditions.
Control rod positions and OPERABILITY of the rod position: indicators are required to be verified on a nominal basis of once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months with more fre-quent verifications required if an automatic monitoring channel is inoperable.
These verification frequencies are adequate for assuring that the applicable LCOs are satisfied.
For Specification 3.1.3.1 ACTIONS b. and c.,
it -is incumbent upon the plant to verify the trippability of the inoperable control rod (s).
Trippability is defined in-Attachment C to a letter dated December 21, 1984, from E. P. Rahe C. O. Thomas NRC).
This may be by verification of a(Westinghouse) tocontrol system failure, usually (el that the failure is associated with the control rod stepping mechanism.
In the event the plant is unable to verify the rod (s) trippability, it must be assumed to be untrippable and thus falls under the requirements of ACTION a.
Assuming a controlled shutdown from 100% RATED THERMAL POWER, this allows approximately 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for this verification.
MILLSTONE UNIT 3 B 3/4 1 4 Amendment No. 60 0027
'f 3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section provide assuran'ce of fuel integrit during Condition I (Normal Operation) and !! (Incidents of Moderate Freqeency events by:
(1) maintaining the minimum DNBR in the core greater than or equa to the design limit during normal operation and in short term transiente, and (2) limiting the fission gas release, fuel pellet temacrature, and chd61ng mechanical properties to within assumed design cr'teria.
In additlon, limiting the peak linear power density durity Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 22f0'F is not exceeded.
The definitions of certain hot channel and peaking factors as used in these specifications are as follows:
F(I)
Heat Flux Hot Channel Factor, is defined as the maximum local heat g
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; and N
F Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of Ag the integral of linear power along the rod with the highest integrated power to the average rod power.
3/4.2.1 AXIAL FLUX DIFFERENCE The limits on AXIAL FLUX DIFFEREN(,E (AFD) assure that the F (Z) upper bound envelope of the F n
n limit specified in the Core Operating Limits Report (COLR) times the normalfled axial peaking factor is not exceeded during either normal operation or in the event of xenon redistribution following power
~~changes, Target flux difference is -determined at equilibrium xenon conditions.~~
i The full length rods may be positioned within tlie core in accordance with their respective insertion limits and should be inserted near their normal i
position for steady state operation at high power levels.
The value of the target flux difference obtained under these conditions divided by the fraction of RATED THERMAL POWER is the target flux difference at RATED THERMAL POWER for the associated core burnup conditions. Target flux differences for other THERMAL POWER levels are obtained by multiplying the RATED THERMAL-POWER value by the appropriate fractional THERMAL POWER level.
The periodic updating of the target flux difference value is necessary to reflect core burnup considerations.
I MILLSTONE UNIT 3 8 3/4 2 1 Amendment No. JJ.60 0028
~-
E POWER DISTRIBUTION LIMITS BASES AXIAL FLUX DIFFERENCE (Continued)
At power levels below APLND, the limits on AFD are defined in the COLR consistent with the Relaxed Axial Offset Control (RAOC) opeating procedure and limits.
These limits were calculated in a manner such that expected operational transients, e.g., load follow operations, would not result in the AFD deviating outside of those limits.
However, in the event such 9 deviation occurs, the short period of time allowed outside of the limit 3 at reduced power levels will not result in significant xenon redistribution such that the envelope of peaking fagrs would change sufficiently to prevent operation in the vicinity of the APL power level.
At power levels greater than APLHD, two modes of operation are permissible:
(1) RAOC, the AFD limit of which are defined in the COLR, and (2) base load operation, which is defined as the maintenance cf the AFD within COLR specgications band about a target value.
The RA00 operat procedure above APL is the same as that defined for operation below APL
However, it is possible when following extended load following maneuvers that the AFD limits may result in restrictions in the maximum allowed power or AFD in order to guarantee operation with Fn(z) less than its limiting value.
To allow operation at the maximum perm'issible power level, the base load operating procedure restricts the indicated AFD to rgatively small-garget band (as specified in the COLR) and power swings (APL 1 power s APL or 100% Rated Thermal Power, whichever is lower).
For base load operation, it is expected that the plant will operate within the target band.
Operation outside of the target band for the short time period allowed will not result in significant xenon redistribution such that the envelope of peaking factors would change sufficiently to prohibit continued operation in the power region defined above.
To assure there is no residual xenon redistribution impact from past operation on tg base load operation, a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months waiting period at a power level above APL and allowed by RA00 is necessary.
During this time period load changes and rod motion are restricted to that allowed by the base load procedure.
After the waiting period, extended base load operation is permissible.
The computer determines the 1 minute average = of each of the OPERA 8LE excore detector outputs and provides an alarm message immediately if the AFD for at least 2 of 4 or 2 of 3 OPERABLE excore channels are:
(1) outside the allowed delta-1 power operating space (for RAOC operation), or (2) outside the allowed delta-1 target band (for base load operation).
These alarms are active when power is greater than (1).50% of RATED THERMAL POWER (for RAOC operation), or MILLSTONE - UNIT 3 B 3/4 2-2 Amendment No. J#,60 0088
~
-.~-
POWER DISTRIBUTION LIMITS BASES AXIAL FLULDIFFERENCE (Continued)
(2)APLHD (for base load operation).
Penalty deviation minutes for base load operation are not accumulated based on the short period of time during which operation outside of the target band is allowed.
3/4.2.2 and 3/4.2.3 r1 EAT F.LVX HOT CHANNEL FACTOR and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR The limits on heat flux hot channel factor, RCS flow rate, and nuclear enthalpy rise hot 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 periodically as specified in Specifications 4.2.2 and 4.2.3.
This periodic surveillance is sufficient to ensure that the limits are maintaint-d provided:
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; b.
Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.6; i
c.
The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained; and d.
The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.
F"H will be maintained within its limitg provided Conditions a. throu A
~~d. above are maintained.~~
The relaxation of F as a function of THERMAL POWER AH allows changes in the radial power shape for all permissible rod insertion limits.
The F"H as calculated in Speciffcations 3.2.3.1 and 3.2.3.2 are used in A
the various accident analyses where FAH influences parameters other than DNBR, e.g.,
peak clad temperature, and tfi'us is the maximum "as measured" value allowed.
MILLSTONE - UNIT 3 8 3/4 2 3 Amendment No. R.60 0028
~.
~
=
- _- ~
m!
I i
POWER DISTRIBUTION LIMITS BASES 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and B.CS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACT @ (Continued) l Margin is maintained between the safety analysis limit DNBR and the design limit DNBR.
This margin is more than sufficient to offset any rod bow penalty and transition core penalty.
The -remaining margin is available for plant design flexibility.
When an F measurement is taken, an allowance for both experimental error n
and manufacturing tolerance must be made.
An allowance of 5% is appropriate i
for a full core map taken with the incore detector flux mapping system and a 3% allowance is appropriate for manufacturing tolerance.
The hot channel factor Ff(z) is measured periodically and increased by a cycle and height dependent power
.4or appropriate to either RAOC or base f
load operation, W(z) or W(z)g to provide assurance that the limit on the hot channel factor, F (z) is met.t,W(z) accounts for the effects of normal opera-4 0
I tion transients aRd was determined from e.xpected power control maneuvers over the full range of burnup conditions in the core.
W(z) accounts for the more restrictive operating limits allowed by base load ophation which result in less severe transient values.
Thew (z)andW(z) actions described above for normal operation are specified in the COLR per Sheificatun 6.9.1.6.
f When RCS flow rate and FN are measured, no additional allowances are necessary prior to comparison hth the limits of the Limiting Condition for Operation.
Measurement errors of 2.4% for four pop flow and 2.8% for three j
loop flow for RCS total flow rate and 4% for F determination of the design DNBR value.
AH have been allowed for in The measurement error for RCS total flow rate is based upon performing a precision heat balance and using the result to calibrate the RCS flow rate indicators.
Potential fouling of the feedwater venturi which might not be detected could bias the result from the precision heet balance in a non-conservative manner.
Therefore, a penalty of 0.1% for undetected fouling of the feedwater venturi will be added if venturis are not inspected and cleaned at least once for 18 months.
Any fouling which might bias.the RCS flow rate measurement greater than 0.1% can be detected by manitoring and trending various 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.
MILLSTONE UNIT 3 8 3/4 2 4 Amendment No. //.,60 0024
l EQWER DISTRIBUTION LIMIT _S_
i BASES HEAT FLUX HOT CHANNEL FACTQR and RC1J10W RATE AND NUCLEAR ENTHALP QL41!iEL FACTOR (Continued)
The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months periodic surveillance of indicated RCS flow is sufficient to detect only flow degradation which could lead to operation outside the accept-able region of operation defined in Specifications 3.2.3.1 and 3.2.3.2.
3/.id 1 00ADRANT POWER TILT RATIO The OVADRANT DOWER TILT RATIO limit assures that the radial power distribution satisfies the design values used in the power capability analysis. Radial power distribution measurements are made during STARTUP testing and periodically during power operation.
The limit of 1.02, at which corrective action is required, provides DNB and linear heat generation rate protection with x y plane power tilts.
A limiting tilt of 1.025 can be tolerated before the margin for uncertainty in F is depleted.
A limit of 1.02 was selected to provide an allowance for the g
u6 certainty associated with the indicated power tilt.
The 2-hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correc-tion of a dropped or misaligned control rod.
In the event such action does not correct the tilt, the margin for uncertainty on F is reinstated by reducingthemaximumallowedpowerby3%foreachpercenthftilt'inexcessof 1.
For purposes of monitcring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the moveable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the QUADRANT POWER !!LT RAT 104 The incore aetector monitoring is done with a full incore flux nap or two sets of four symmetric thimbles.
The two sets of four symmetric thimbles is a unique set of eight detector locations.
These-locations are C 8, E 5 E II, H 3, H 13, L 5, L ll, N 8.
}/4.2.5 DNB PARAMETERS t
The limits on the DND related. parameters assure that each of the parameters are 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 greater than the design limit throughout each analyzed transient.
The indicated T,yg value of 591.l'F (four loop MILLSTONE - UNIT 3 B 3/4 2-5 Amendment No. 27, 5,60 0026
. _ _.., _ _. ~.. _
POWER DISTRIBUTION LIMITS BASES DNB PARAMETERS (Continued) operatiois) or 583.3'F (three loops operating) and the indicated pressurizer pressure value is 2218 psia (four loop or three loop operation The calculated values of the DNB related parameters will be an average).
o ~ the indicated values for the operable channels.
The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 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.
Measurement uncertainties have been accounted for in determining the parameter limits.
MILLSTONE - UNIT 3 B 3/4_2 6 Amendment No U,60 00t4
q 3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The p'.',t is designed to operate in MODES 1 and 2 with three or four reactor coC. ant loops in operation and maintain DNBR greater than the design limit during all normal operations and anticipated transients.
With less than the required reactor coolant loops in operation this specification requires that the plant be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
In MODE 3, three reactor coolant loops, and is Mode 4, two reactor coolant loops provide sufficient heat removal capablitt:y for removing core decay heat even in the event of a bank withdrawal accident; however, a single reactor coolant loop provides sufficient heat removal capacity if a bank withdrawal accident can be prevented, i.e., by opening the Reactor Trip System breakers.
In MODE 4, and in MODE 5 with reactor coolant loops filled, a single reactor coolant loop or RHR loop provides sufficient heat removal capability for removing decay heat; but sin le failure considerations require that at leasttwoloops(eitherRHRorRCS be OPERABLE.
In MODE 5 with reactor coolant loops not filled, a single RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations, and the unavailability of the steam generators as a heat removing component, require that at least two RHR loops be OPERABLE.
The locking closed of the required valves in Mode 5 (with the loops not filled) will preclude the possibility of uncontrolled baron dilution of the Reactor Coolant System by preventing flow to the RCS of unborated water.
The operation of one reactor coolant pump adequate flow to ensure mixing, prevent strat(fication and produce gradualRC reactivity changes during boron concentrati o reductions in the Reactor Coolant System.
The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control.
The rattrictions on starting an RCP with one or more RCS cold legs less than or equal to 350'F are provided to prevent RCS pressure transients, caused by energy additions from the Secondary Coolant-System, which could exceed the limits of Appendix G to 10 CFR Part 50.
The RCS will be protected against overpressure -transients and will not exceed the limits of Appen61x G by l
either:
(1) restricting the water volume in the pressurizer and thereby providing a volume for the reactor coolant to expand into, or (2) by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 50'F above each of the RCS cold leg temperatures.
MILLSTONE UNIT 3 8 3/4 4 1 Amendment No. 7,60 1
00t9
--_----_,-----w--.-n,--
,e---r--------ww-
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l 3/4.4 REACTOR COOLANT SYSTEM i
i BASES (Continuedi l
The requirement to :.aintain the isolated loop stop valves shut with power removed ensures that no reactivity addition to the core could occur due to the i
startup of an isolated loop.
Verification of the boron concentration in an i
idle loop prior to opening the stop valves provides a reassurance of the i
adequacy of the boron concentration in the isolated loop.
The 2600 ppm is sufficient to bound shutdown margin requirements and provide for boron r
concentration measurement uncertainty betwen the loop and the RWST.
Draining i
and refilling the isolated loop within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to opening its stop valves ensures adequate mixing of the coolant in this loop and prevents any reactivity effects due to boron concentration stratifications.
P t
i I
i l
J f
t I
MILLSTONE - UNIT 3 83/441a Amendment No.],60 l 0029
mi 2/19 REFUELING OPERATIONS BASES 3/4.9.1 BORON CONCENTRATION The limitations on reactivity conditions during REFUELING ensure thatt (1) the reactor will remain suberitical during CORE ALTERATIONS, and uniform boron concentration is maintained for reactivity control in the w(2) a ater volume having direct access to the reactor vessel.
The value of 0.95 or less for K a
1% Ak/k conservative allowance for uncertainties.
Similah, includes the boron concentration value of 2600 ppm or greater includes a conservative uncertainty allowance of 50 ppm boron.
The 2600 ppm provides for boron concentration measurement uncertainty between the spent fuel pool and the RWST.
The locking closed of the required valves during refueling operations precludes the possibility of uncontrolled boron dilution of the filled portion of the RCS.
This action prevents flow to the RCS of unborated water by closing flow paths from sources of unborated water.
3/4.9.1.2 Boron Concentration in Snent Fuel Pool The limitations of this specification ensure that in the event of a fuel assembly handling accident involving either a misplaced or dropped fuel assembly, the Kggy of the spent fuel storage racks will remain less than or equal to.95.
3/4.9.2 INSTRUMENTATION The OPERABILITY of the Source Range Neutron Flux Monitors ensures that redundant monitoring ca) ability is available to detect changes in the reactivity condition of tie core.
3/4.9.3 DECAY TIME The minimum requirement for reactor subcriticality prior to movement of irradiated fuel assemblies in the reactor vessel ensures that sufficient time has elapsed to allow the radioactive decay of the short lived fission products. This decay time is consistent with the a:sumptions used in the safety analyses.
3/4.9.4 CONTAINMENT BUILDING PENETRATIORS The requirements on containment building penetration closure and OPERABILITY ensure that a release of radioactive material within containment will be restricted from leakage to the environment.
The OPERABILITY and closure restrictions are sufficient to restrict radioactive material release from a fuel element rupture based upon the lack of containment pressurization potential while in the REFUEllHG MODE.
3/4.9.5 COMMUNICATIONS The requirement for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the i
)
facility status or core reactivity conditions during CORE ALTERATIONS.
i MILLSTONE - UNIT 3 4 ?/4 9 1 Amendment No. #.60 0030
ceg DESIGN FEATURES C
5,6 FUEL STORAGE CRITICALITY 5.6.1.1 The spent fuel storage racks are designed and shall be maintained with:
a.
Ak unb8Ntedwater. equivalent to less than or equal to 0.95 when flooded with b.
A nominal 10.35 inch center-to center distance between fuel I
assemblies placed in the storage racks, Fuel assemblies stored in Region 1 of the spent fuel pool may have a c.
i maximum nominal fuel enrichment of up to 5.0 weight percent U Region I is designed to permit storage of fuel in a 3 out-$N4 array with the 4th storage location blocked as shown in Figure 3.9-2, d.
Fuel assemblies stored in Region II of the spent fuci pool may have a maximum nominal fuel enrichment of up to 5.0 weight percent, conditional upon compliance with Figure 3.9-1 to ensure that the design burnup of the fuel has been sustained.
DfBlNME l
i 5.6.2 The spent fuel storage pool is designed and shall be maintained tt prevent inadvertent draining of the pool below elevation 45 feet.
CAPACITY 5.6.3 The spent fuel storage pool contains 756 storage locations of which a maximum of 100 locations will be blocked.
l 5.7 COMP 0NENT CYCLIC OR TRANSIENT llMIT 5.7.1 The components identified in Table 5.71 are designed and shall be maintained within the cyclic or transient limits of Table 5.7-1.
l MILLSTONE - UNIT 3 56 Amendment No. Af,60
-.--.s,-
,q ADMINISTRATIVE CONTROLS
.SEMIAN!iVAL,RADI0 ACTIVE EFFLUENT RELEASE REPORT
6.9.1.4 Routine Semiannual Radioact te Effluent Release' Reports covering the operation of the unit during the previous 6 months of operation shall be i
submhted within 60 days after January 1 and July 1 of each year, A supplemental report containing dose assessments for the previous year shall be submitted annually within 90 days after January 1.
The report shall include that information delineated in the REH0D3.
1 Any changes to the REMODCM shall be submitted in the Semiannual Radioac-tive Effluent Release Report.
MONTHLY OPERAT_ING REPORTS 6.9.1.5 Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the U.S. Nuclear Regulatory Comission, Document Control Desk, Washington, D.C.
20555, one copy to the Regional Administrator Region 1, and one copy to the NRC Resident inspector, no later than the 15th of each month following the calendar month covered by the report.
CORE OPERATING llMITS REPOR1 6.9.1.6.a Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:
1.
Moderator Temperature Coefficient BOL and EOL limits and 300 ppm surveillance limit for Specification 3/4.1.1.3, 2.
Shutdown Rod Insertion Limit for Specification 3/4.1.3.5, 3.
Control Rod Insertion Limits for Specification 3/4.1.3.6, 4.
Axial Flux Difference Limits, target band, and APLND tions 3/4.2.1.1 and 3/4.2.1.2, for Specifica-1 5.
Heat Flux Hot Channel Factor, K(z), W(z), APLND and W(z)BL fI Specifications 3/4.2.2.1 and 3/4.2.2.2.
I 6.
Nuclear Enthalpy Rise Hot Channel Factor, Power Factor Multiplier for Specification 3/4.2.3.
~~A si'ngle submittal may be made for a multiple unit station.~~
The submittal should combine those sections that are common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.
1 Mll.LSTONE - UNIT 3 6-21 Amendment No. 4, A7, S,60 i
0031
Othg ;
6.9.1.6.b The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in:
~
1.
WCAP 9272 P A, " WESTINGHOUSE RELOAD SAFETY EVALVATION METHODOLOGY,"
July 1985 (W Proprietary). (M&thodology for Specifications 3.1.1.3 -
Moderator Temperature Coefficient, 3.1.3.5--Shutdown Bank Insertion Limit, 3.1.3.6 -Control 8ank 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.)
2.
WCAP 8385,
" Power Distribution Control and Load _
Following Procedures - Topical Report," September 1981 (W Proprietary).
3.
T. M. Anderson to K. Kniel (Chief of Core Performance Branch, NRC),
January 31, 1980 -
~~Attachment:~~
Operation and Safety Analysis Aspects of an Improved Load Follow Package.
4.
NUREG 800, Standard Review Plan, U.S. Nuclear Regulatory Comission, Sect. ion 4.3, Nuclear Design, July 1981 Branch Technical Position CPB 4.3 1, Westinghouse Constant Axial Offset Control (CAOC),
Revision 2, July 1981.
S.
WCAP-10216 P A,
' RELAXATION OF CONSTANT AXIAL OFFSET CONTROL F SURVEILLANCE TECHNICAL SPECIFICATION," June 1983 (W Proprietary (Methodology for Specifications 3.2.1--Axial Flux Difference (Relaxed Axial Offset Control) and 3.2.2-Heat Flux Hot Channel Factor (W(2) surveillance requirements for Fg ethodology).)
M 6.
9 CAP-9561 P A, ADO. 3, Rev.1, "BART A 1:
A COMPUTER CODE FOR THE BEST ESTIMATE ANALYSIS OF REFLOOD TRANSIENTS-SPECIAL REP THIMBLE MODELING W ECCS EVALVATION MODEL," July 1986 (W Proprie-tary).
(Methodology for Specification 3.2.2-Heat Flux Hot Channel Factor.)
7.
WCAP 10266 P A, Rev. 2. "THE 1981 VERSION OF WESTINGHOUSE CVALUATION MODEL USING BASH CODE," March 1967 (W Proprietary).
l for Specification 3.2.2--Heat Flux Hot Channel Factor.) (Hethodology 1
8.
WCAP 11946,
" Safety Evaluation Supporting a More Negative EOL Moderator Temperature Coefficient Technical Specification for the Millstone Nuclear Power Station Unit 3," September 1988 (W Proprie-i tary),
i 6.9.1.6.c The core o erating limits shall be determined so that all applica-ble limits (e.g.
fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.
6.9.1.6.d The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions cycle,pplements thereto, or su shall be provided upon issuance, for each reload to the NRC Document Control Desk with copies to the Regional Adminis-trator and Res1&nt Inspector.
I gSTONE-UNIT 3 6-214
-Amendment No. g,60
- - - _

ML20070K460

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