Text

Amend 12 to License NPF-49,revising Several Tech Specs Sections to Accommodate Cycle 2 Reload,Installation of Resistance Temp Detectors in RCS & Restart of Isolated Reactor Coolant Loop
	ML20149C916
Person / Time
Site:	Millstone
Issue date:	01/20/1988
From:	Stolz J; Office of Nuclear Reactor Regulation
To:	; Northeast Nuclear Energy Co (NNECO)
Shared Package
ML20149C921	List: ML20149C916; ML20149C951;
References
	NPF-49-A-012 NUDOCS 8802090339
	Download: ML20149C916 (53)
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UNITED STATES

'g j

'g NUCLE AR REGULATORY COMMISSION 3

5.8 W ASHING TON. D. C. 205$5

,e NORTHEAST NUCLEAR ENERGY COMPANY, ET AL.*

DOCKET NO. 50-423 MILLSTONE NUCLEAR POWER STATION, UNIT NO. 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 12 License No. NPF-49 1.

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

I A.

The application for amendment by Northeast Nuclear Energy Company, et al. (the licensee) dated September 9, 1987, and supplemental 4

i letters dated September 9, 1987, and September 30, 1987 comply i

with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the tommission; 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 Commission's regulations; i

D.

The issuance of this amendment will not be inimical to the common

]

defense and security or to the health and safety of the public; and t

i E.

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

6

Northeast Nuclear Energy Company is authorized to act as agent and represent-ative for the following Owners:

Central Maine Power Company, Central Vermont Public Service Corporation, Chicopee Municipal Lighting Plant, City of i

Burlington, Vermont, Connecticut Municipal Electric Light Company, Massachusetts Municipal Wholesale Electric Company, Montaup Electric Company, New England Power Company, The Village of Lyndonville Electric Department Western Massachusetts Electric Company, and Vermont Electric Generation and

]

Transmission Cooperativa, Inc., and has exclusive responsibility and control i

i over the physical construction, operation and maintenance of the facility.

8802090339 880120' PDR ADOCK 05000423 P

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2.

Accordingly, the license is amended by chenges to the Technical Specifications as indicated in the attachment to this license amendraent, i

and paragraph 2.C.(2) of Facility Operating Licer.se No. NPF-49 is hereby i

amended to read as follows:

i 4

j' (2) Technical Specifications The Technical Specifications contained in Appendix A as revised through Aniendment Nc 12

, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto are hereby incorporated in the license.

The licensee shall operate the w

facility in accordance with the Technical Specifications and the j

Environmental Protection Plan.

3.

This license amendment is effective as of the date of its issuance.

[

6 FOR THE NUCLEAR REGULATORY COMMISSION i

[

'f i

Jo n. Stolz, Direc r l

Pr ject Directorat

-4 vision of Reactor Projects I/II i

1 j

Office of Nuclear Reactor Regulation

)

Attachment:

l Changes to the Technical i

4 Specifications Date of Issuance: January 20, 1988 i

4 i

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e ATTACHMENT TO LICENSE AMENDMENT NO. 12 FACILTIY OPERATING LICENSE NO. NPF-49 DOCKET NO. 50-423 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages.

The revised pages are identified b contain vertical lines indicating the areas of change. y amendment number and Remove Insert 25 2-5 2-6 2-6 2-8 2-8 2-9 2-9 2-11 B2-5 2-11 3/4 1-4 B2-5 3/4 1-11 3/4 1-4 3/4 1-12 3/4 1-11 3/4 2-15 3/4 1-12 3/4 2-18 3/4 2-15 3/4 2-24 3/4 2-18 3/4 3-8 3/4 2-24 3/4 3-10 3/4 3-8 3/4 3-10 3/4 3-14 3/4 3-10a 3/4 3-28 3/4 3-14 3/4 3-30 3/4 3-28 3/4 3-34 3/4 3-30 3/4 4-8 3/4 3-34 3/4 5-1 3/4 4-8 3/4 5 9 3/4 5-1 3/4 6-14 3/4 5-9 3/4 9-1 3/4 6-14 3/4 9-1 B3/4 1-3 3/4 9-la B3/4 1-3 83/4 2-5 B3/4 1-3a B3/4 2-6 B3/4 2 5 83/4 2-7 B3/4 2 6 83/4 9-1 B3/4 2-7 5-5 B3/4 9-1 5-5

,T'.alLE 2.2-1 REACTOR TRIP SY5 REM INSTRUMENTATION TRIP SETPOINTS x

SENSOR

~.

E TOTAL ERROR FUNCTIONAL UNIT ALLOWANCE (TA)

~

(S)

TRIP SETPOINT ALLOWABLE VALUE Z

O M

1.

Manual Reactor Trip N.A.

N.A.

c:

2.

Power Range, Neutron Flux 5

]

a.

High Setpoint

1) Four Loops Operating 7.5 4.56 0

di109% of RTP" dh111.1% of RTP"

2) Three Loops Operating 7.5 4.56 0

db80% of RTP

d6 82.1% of RTPee b.

Low Setpoint 8.3 4.56 O

fE 25% or RTP

fi27.1% of RTP" 3

Power Range, Neutron Flux, 1.6 0.5 0

6: 5% of RTP

fi 6.3% or RTP

High Positive Rate with a time a time constant contant 2 2 seconds 2*2 seconds 4.

Power Range, Neutron Flux, 1.6 0.5 0

f= 5% o r RTP" with fL6.3% or RTP" with High Negative Rate a time constant a time constant 2 2 seconds 25 2 seconds a

5.

Intermediate Range, 17.0 8.41 0

f 25% of RTP

6 30.9% of RTP

Neutron Flux 6.

Source Range, Neutron Flux 17.0 10.01 0

fL10+5 ep3 d; 1,4 x to+5 ep3 4

7.

Overtemperature A T

$g a.

Four Loops Operating 8.3 5.76 1.6T + 1.17 See Note 1 See Note 2 j[

S (Temp + Press) c' b.

Three Loops Operating 12.0 5.77 1.73 + 1.17 See Note 1 See Note 2 jy (Temp + Press) 8.

Overpower di T 4.8 1.22 1.67 See Note 3 See Note 4

' Loop oesign riow = 94,600 gpm (Four Loops Operating); 99,600 (Three Loops Operating) seRTP = RATED THERMAL POWER

TABLE 2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 3

r SENSOR

[*,

TOTAL ERROR FUNCTIONAL UNIT ALLOWANCE (TA)

Z (S)

TRIP SETPOINT ALLOWABLE VALUE x

9.

Pressurizer Pressure-Low 5.0 1.77 33 2h 1900 psia 2t 1890 psia E

10.

Pressurizer Pressure-High 5.0 1.77 3.3 fi 2385 psia di 2395 psia U

u, 11.

Pressurizer Water Level-High 8.0 5.13 2.7 fi 895 or di 90.7% or-instrument spv1 instrument span 12.

Reactor Coolant Flow-Low 2.5 1.52 0.78 25 90% or loop 25 89.1 % or loop design flow' design flow'

13. Steam Generator Water 20.5 18.98 1.75 2r 23.5% or narrow 25 22.6% of narrow Level Low-Low range instrument range instrument span span 14.

General Warning Alarm N.A.

N.A.

7 15.

Low Shart Speed - Reactor 3.8 0.5 0

25 97.8% or rated 25 94.6% or rated Coolant Pumps speed speed 16.

Turbine Trip a.

Low Fluid Oil Pressure N.A.

N.A.

25 500 psig it"450 psig b.

Turbine Stop Valve N.A.

N.A.

Et 1% open

'E.1% open Closure 17.

Sarety Injection Input N.A.

N.A.

37 from ESF 5.

o k

- RTP = RATED THERMAL POWER M

TABLE 2.2-1 (Continued)

~

3a p

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS o

SENSOR 5

TOTAL ERROR e

FUNCTIONAL UNIT ALLOWANCE (TA)

Z g)

TRIP SETPOINT ALLOWABLE VALUE E

18.

Reactor Trip System Interlocks w

a.

Intermediate Range N.A.

N.A.

3 1 x 10 O amp

> 6 x 10 18 amp Neutron Flux, P-6 b.

Low Power Reactor Trips Block, P-7

1) P-10 input N.A.

N.A.

N. A.

$ 10% of RTP**

1 12.1% of RTP**

2) P-13 input N.A.

N.A.

5 10% RTP** Turbine $ 12.1% RTP** Turbine Impulse' Pressure Impulse Pressure "4

Equivalent Equivalent c.

Power Range Neutron Flux, P-8

1) Four Loops Operating N.A.

N.A.

$ 37.5% of RTP**

$ 39.6% of RTP**

2) Three Loops Operating N.A.

N.A.

$ 37.5% of RTP**

$ 39.6% of RTP**

d.

Power Range Neutron N.A.

N.A.

$ 51% of RTP**

$ 53.1% of RTP**

Flux, P-9 e.

Power Range Neutron N.A.

N.A.

3 10% of RTP**

3 7.9% of RTP**

Flux, P-10 19.

Reactor Trip Breakers N.A.

N.A.

N.A N.A.

N.A.

20.

Automatic Trip and Interlock N.A.

N.A.

Logic 21.

Three Loop Operation N.A.

N.A.

N.A Bypass Circuitry

- RTP = RATED THERMAL POVER 1

3 TABLE 2.2-1 (Continued) o M

TABLE NOTATIONS i

e NOTE 1: OVERTEMPERATURE fA T 5

1 1

( --* **S} [T (1 + t AT ((1 + t,5)

5) < ATo [Ka - K

.S) - T'] + K (P - P') - f (at))

3 2 (1 + 1 5) 1+T 5) 1+T 3

2 3

= Measured A T by Reactor Coolant System Instrumentation; Where:

AT 1 + tsS Lead-lag compensator on measured AT;

=

5 1+12 Time constants utilized in lead-lag compensator for A T,

'rt = 12 s,

=

13,12 T2 = 3 s; 1

Lag compensator on measured AT;

=

1+t53 Time constants utilized in the lag compensator for A T,

'T3 = 0 s;

=

T3

= Indicated A T at RATED TIIERMAL POWER; ai O

1.08 (Four Loops Oparating); 1.01 (Three Loops Operating);

K,

=

g K,

0.01313; I

=

.i m

I E

1*T S The function generated by the lead-lag compensator for T A

=

g 1+1 5

- 8 5

dynamic compensation; e

T4, is Time. constants utilized in the lead-lag compensator for T,,g, "Ca = 33 s, T5 = 4 s; Average temperature, OF; y

=

I i

1 Lag compensator on measured T I + ts5 avg; i

is

Time constant utilized in the meanured T lag compensator, T6=0s, avg i

l

3p 4

O TA3LE 2.2-1 (Continued)

E TABLE NOTATIONS (Continued)

U t.,

NOTE 1:

(Continued)

T' 6 587.10F (Nominal T at RATED TilERMAL POWER);

K

=

0.000663/ psi; 3

P

=

Pressurizer pressure, psta; P'

=

2250 psia (Nominal RCS operating pressure);

Laplace transform operator, s-l S

=

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

the power-range neutron ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:

u E

(1)

For q

-q between -30% and + 105, r AI = 0, where q and q are percent RATED THERMkLPONERinthetopandbottomhd(veso)fthecorerespectivNly,ando l

+q is total THERMAL POWER in pc. cent of RATED THERMAL POWER; (2) For each percent that the magnitude of q

-q exceeds -30%, the A T Trip Setpoint g

shall be automatically reduced by 3.6% of its value at RATED TilERMAL POWER; and a,

5 (3) For each percent that the magnitude of q

-g ex e ds +10%, the A T Trip Setpoint 5

shall be automatically reduced by 2.0% ob its value at RATED THERMAL POWER.

b E

The channel's maximum Trip Setpoln't shall not exceed its computed Trip Setpoint by more NOTE 2:

N than 2.1% A T span (Four Loop Operation); 3.6% A T span (Three Loop Operation).

TABLE 2.2-1 (Continued) xF TABLE NOTATIONS (Continued)

G

$5 NOTE 3: OVERPOWER AT "U

- T9 - f (AI))

O 5

E 2

1 + IaS 0

1+T 5 (1 + Tc5 1

Tc5

+

7 t.d Where:

AT As defined in Note 1,

=

f{

As defined in Note 1.,

=

As defined in Note 1,

=

ti. T2 1

As defined in Note 1,

=

1+T 5 7

3 5

13 As defined in Note 1,

=

AT, As defined in Note 1,

=

4 1.09, K

=

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

=

temperature,

{7 M e funcd on genera W h %e ra$ed ag c g ensator for Tavg @namic

=

3 7

compensation, T7 Time constants utilized in the rate-lag compensator for T

,T7 10 s.

=

I As defined in Note 1,

=

1 + Tc5 As defined in Note 1,

=

is t

- ~.

- - _ =. _. -

t

'o x

5 Gg M

TABLE 2.2-1 (Continued) ex TABLE NOTATIONS (Continued) w NOTE 3:

(Continued) 0.00129/0F for T > T" and K6 = 0 for T 6. T",

K

=

6 As defined in Note 1, '

T

=

T"

= Indicated T at RATED THERMAL POWER (Calibration temperature for A T instru*mEbtatiors.

6 587.10F),

S

=

As defined in Note 1, and T ( A I) 2

= 0 for all A I.

NOTE 4:

The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 2.8% A T span.

g M

N i

4

s 4

LIMITING SAFETY SYSTEM SETTINGS BASES Intermediate and Source Range, Neut: en Flux The Intermediate and Source Renge, Neutron Flux trips provide core protection during reactor startup to citigate the consequences of an uncon-trolled rod cluster control assembly bank withdrawal from a suberitical 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 Intercediate Range channels will initiate a Reactor trip at a current level equivalent to approximately 25% of RATED THERMAL POWER unless canually blocked when P-10 becoces active.

No credit was taken for operation of the trips associated with either the Inter =ediate 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.

Overte:perature 4>T The Overte:perature 4LT trip provides core protection to prevent DR3 for all combinations of pres'sure, power, coolant temperature, and axial power distribution, provided that the transient is slow with respect to piping transit delays from the core to the te=perature 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 te:perature induced changes in density and heat capacity of water and includes dynamic co:pensation for piping delays from the co're to the loop te:perature detectors, (2) pressurizer pressure, and (3) axial power distrib6 tion.

With normal axial power distribution, this Reactor trip limit is always below the core Safety Limit as shown in Figure 2.1-1. If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the Reactor trip is automatically reduced according to the notations in Table 2.2-1.

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 Overteeperature 4L T channels, reducing the Power Range Neutron Plux High setpoint to a value just above the three loop maxi =um permissible power level, and resetting the P-B setpoint to its three loop I

value.

These codifications have been chosen so that, in three loop operation, each co:ponent of the Reactor Trip System performs its normal four loop function, prevents operation outside the safety limit curves, and prevents the DRSR fro: going below 1.30 during normal operational end anticipated transients.

Overpoweras T The Overpower Jk T trip provides assurance of fuel integrity (e.g., no fuel pellet melting and less than 15 cladding strain) under all possible overpower conditions, limits the required range for Overtemperature 45 T MILLSTONE - UNIT 3 B 2-5 Amendment No.12

-~

1 1

LIMfTING SAFETY SYSTEM SETT1NGS BASES I

I trip, and provides a backup to the High Neutron Flux trip.

The Setpoint is automatically varied with:. (1) coolant temperature to correct for tempera-ture induced changes in density and heat capacity of water, and (2) rate of change of temperature for dynamic compensation for piping delays from the ce a to the loop temperature detectors, to ensure that the allowable heat genera-tion rate (kW/ft) is not exceeded.

The Overpower aT trip provides protection to mitigate the consequences of various size steam breaks as reported in WCAP-9226, "Reactor Core Response to Excessive Secondary Steam Releases."

Pressurizer Pressure In each of the pressurizer pressure channels, there are two independent bistacles, each with its own trip setting to provide for a High and Lo Pressure trip thus limiting the pressure range in which reactor operation is permitted.

The Low Setpoint trip protects against lo-pressure which could lead to DNB bj tripping the reactor in the event of a loss of reactor coolant pressure.

On decreasing po er the Low Setpoint trip is automatically blocked by P-7 (a po er level of approximately 10% of RATED THERMAL POWER with turbine impulse cha-ber pressure at approximately 10% of full power equivalent); and on increasing po er, automatically reinstated by P-7.

The High Setpoint trip functions in conjunction with the pressurizer re'ief and safety valves to protect the Reactor Coolant System against system overpressure.

Pressuri2er Water Level The Pressurizer Water Level High trip is provided to prevent water relief through the pressurizer safety valves.

On decreasing power the Pressurizer High Water Level trip is automatically blocked by P-7 (a power level of approxi-mately 10% of RATED THERMAL POWER with a turbine impulse chamber pressure at approximately 10% of full power equivalent); and on increasing power, auto-matically reinstated by P-7.

Reactor Coolant Flow The Reactor Coolant Flo Low trip provides core protection to prevent DNS by mitigating the consequences of a loss of flow resulting from the loss of one or more reactor coolant pumps.

On increasing power above P-7 (a power level of approximately 10% of RATED THERMAL POWER or a turbine impulse chamber pressure at approximately 10%

of full power equivalent), an automatic Reactor trip will occur if the flow in more than one loop drops below 90% of nominal full loop flow.

Above P-8 (a power level of approximately 38% of RATED THERMAL POWER) an automatic Reactor trip will occur if the flow in any single loop drops below 90% of nominal full loop flow.

Conversely, on decreasing power between P-8 and the P-7 an automatic Reactor trip will occur on low reactor ecolant flow in more than i

one loop and below P-7 the trip function is automatically blocked, i

MILLSTONE - UNIT 3 8 2-6 a

REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN - T,yg LESS THAN OR EQUAL TO 200 F LIMITING CONDITION FOR OPERATION 3.1.1.2 The SHUTOOWN MARGIN shall be greater than or equal to 1.6% ok/k.

APPLICABILITY:

MODE 5.

ACTION:

With the SHUT 00WN HARGIN less than 1.6% ok/k, 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 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1.6% ak/k:

Within 1 hou'r 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 SHUTDOWN MARGIN shall bt verified acceptable with an increased allowsnce 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 b;Jrnup based on gross thermal energy generation,

=,

5)

Xenon concentration, and 6)

Samarium concentration.

MILLSTONE - UNIT 3 3/4 1-3

o REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION

~3 1.1.3 The moderator temperature coefficient (MTC) shall be:

a.

Less positive than +0.5 x 10-4 46k/k/0F for the all rods withdrawn, beginning of cycle life (BOL) condition for power levels up to 70%

RATED THERMAL POWER with a linear raep to O Abk/k/0F at 100% RATED THERMAL POWER.

b.

Less negative than -4.0 x 10-4 Ak k/k/0F for the all rods withdrawn, end of cycle life (EOL), RATED THERMAL POWER condition.

APPLICABILITY: Specification 3 1.1 3a. - MODES 1 and 28 only".

Specification 3.1.1 3b. - MODES 1, 2, and 3 only".

ACTION:

a.

With the MTC more positive than the limit of Specification 3.1.1.32.

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

~

1.

Contro1 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 HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . These withdrawal limits shall be in addition to the insertion limits of Specification 3 1 3 6; 2.

The control rods are mnintained w[ thin the withdrawal limits established above unt 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.the Commission, pursuant to Specification 6.9.2, within 10 days, describing the value of the measured MTC, the interim control rod withdrawal limits, and the predicted average core burnup necessary for restoring the positive MTC to within its limit for the all rods withdrawn condition.

b.

With the MTC more negative than the limit of Specification 3.1.1.3b.

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

'With K,pp greater than or equal to 1.

esSee Special Test Exceptions Specification 310 3 MILLSTONE - UNIT 3 3/4 1-4 Amendment No.12

5 REACTIVITY CONTi:0L SYSTEMS BORATED WATER SOURCE - SHUTLOWN LIMITING CONDITION FOR OPERATION 3.1.2.5 As a minimum, one of the'following borated wat'er sources shall be OPERABLE:

a.

A Boric Acid Storage System with:

1)

A minimum contained borated water volume of 6700 gallons, 2)

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

A minimum solution temperature of 670F.

b.

The refueling water storagt tank (RWST) with:

1)

A minimum contained borated water volume of 250,000 gallons, 2)

A =inimum boron concentration of-2300 ppm, and 3)

A minimu$ solution temperature of 400F.

_ APPLICABILITY: MODES 5 and 6.

ACTION:

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

SURVEILLANCE REQUIREMENTS 4.1.2.5 The above required berated water source shall be demonstrated OPERABLE:

a.

At least once per 7 days by:

1)

Verifying the boron concentration of the water, 2)

Verifying the contained borated water volume, and 3)

Verifying the Boric Acid Transfer Pump Room temperature and the l

boric acid storage tank solution temperature when it is the source of berated 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 350F.

MILLSTONE - UNIT 3 3/4 1-11 Amendment No.12

i e

REACTIVITY CONTROL SYSTEP S BORATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 3 1.2.6 As a minimum the following borated water source (s) shall be OPERABLE l

as required by Specification 3 1.2.2:

i a.

A Boric Acid Storage System with:

1)

A minimum contained borated water volume of 23,620 gallons, 2)

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

A minimum solution temperature of 670F.

b.

The refueling water storage tank (RWST) with:

1)

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

2)

A boron concentration between 2300 and 2600 ppe, 3)

A minimum solution temperature of 400F, and 4)

A maximus solution temperature of 500F.

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

ACTION:

With the Boric Acid Storage System inopsrable, restore,the system to a.

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 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 1.6% A k/k at 2000F; restore the Boric Acid Storage System to OPERABLE status within the next 7 days or be in COLD SHUTDOWN 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 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within the next 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 .

MILLSTONE - UNIT 3 3/4 1-12 Amendment No.12

POWER DISTRIBUTION LIMITS 3/4.2.3 RCS PLOW RATE AND NUCLEAR'ENTHALPY RISE HOT CHANNEL FACTOR FOUR LOOPS OPERATING LIMITING CONDITION FOR OPERATION The indicated Reactor Coolant System (RCS) total flow rate and ([g 3231 shall be maintained as follows:

a.

RCS total flow rate at 385,210 gpm, and N dE1.49(I.0+0.3(1.0-P))

b.

F g Where:

THERMAL POWER g) p,

RATED THERMAL POWER 2)

F easured values of F btained by using the AH =ble in-core detectors kH mova o obtain a power The measured value of F'"Ihto distribution cap.

.should be used since Specification 3 2 3.1b.

takes consideration a measurement uncertainty of 4% for incere meas'rement, and u

3)

The measured value of RCS total flow rate shall be used since uncertainties of 1.8% for flow measurement have been included in Specification 3.2.3.1a.

APPLICABILITY: MODE 1.

ACTION:

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

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 F to within the g

above limits, or i

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 .

)

l MILLSTONE - UNIT 3 3/4 2-15 Amendment No. 12 1

O POWER DISTRIBUTION' LIMITS i

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 limits, verify through incore flux mapping and RCS total flow rate that F and H

RCS total flow rate are restored to within the above limits, or reduce THcRMAL 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 , c.

Identify and correct the cause of the out-of-limit condition prior to increasing T?ERMAL 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 rate are H

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 REQUIREMENTS 4.2.3.1.1 The provisions of Specification 4.0.4 are not applichble.

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

acceptable range:

a.

Prior to operation above 75% of RATED THERMAL POWEit af ter each fuel 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 value of F H, 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 CALIBRATION at least once por 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 2-16

. POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) 4.2.3.1.5 The RCS total flow rate shall be determined by precision heat balance measurement at least once per 18 months.

Within 7 days prior to performing the precision heat bala,nce, the instrumentation used for deter-mination of steam pressure, feedwater pressure, feedwater temperature, and feedwater venturi SP 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.

d S

0 l

MILLSTONE - UNIT 3 3/4 2-17

o POWER DISTRIBUTION LIMITS RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR THREE LOOPS OPERATING LIMITING CONDITION FOR OPERATION 3.2 3.2 The indicated Reactor Coolant System (RCS) total flow rate and FN shall be maintained as follows:

g 1

a.

RCS total flow rate & 304,780 gpm, and b.

F 61351 {.0+0.43(1.0-Ph 3g Where:

THERMAL POWER RATED THERMAL POWER

2) F H=

Measured values of F a ne AH y us ng We movable incore detectors to obtain a power distribution map.

The' measured value of F should be used since Speci-3g fication 3.2 3.2b. takes into consideration a measure-

tent uncertainty of 45 for incore measurement, and 3)

The measured value of RCS total flow rate shall be used since uncertainties of 2.0% for flow measurement have been included in Specification 3 2 3.2a.

APPLICABILITY: MODE 1.

ACTION:

With the RCS total flow rate or F I

operation:

4H utside the region of acceptable 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 F limits, or AH 2.

Reduce THERMAL POWER to less than 32% of RATED THERMAL POW and reduce the Power Range Neutron Flux - High Trip Setpoint to less than or equal to 37% 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 2-18 bendmentNo.12

\\

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 THERHAL 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 .

1 MILLSTONE - UNIT 3 3/4 2-23

x TABIE 3.2-1 5

C DNB PARAETERS a

N a

LIMITS c

'd Three Loops in Opera-Four Loops in tion & Loop Stop PARAEER Operation Valves Closed Indicated Reactor Coolant System T,y f 591.20F

{- 583.fl0F Indicated Pressurizer Pressure N 2226 psia'

> 2226 psia

R c-Y "v-5 S

11

$e F

Y

' Limit not applicable during either a TIERMAL POWER ramp in excess or Si or RAED TIERMAL,

POER per minute or a THERMAL POER step in excess or 10% or RATED TIERMAL POER.

4 TABLE 3.3-1 (Continued)

ACTION STATEMENTS (Continued)

ACTION 9 - With a channel associated with an operating loop inoperable, restore the inoperable channel to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 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 .

One channel associated with an operating loop may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.1.1.

ACTION 10 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.1.1, provided the other channel is OPERABLE.

ACTION 11 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open the Reactor Trip System breakers within the next hour.

ACTION 12 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

The -inoperable channel is placed in the tripped condition a.

within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and b.

When the Minimum Channels OPERABLE requirement is met, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of the Turbine Control Valves.

ACTION 13 - With one of the diverse trip features (undervoltage or shunt trip attachments) inoperable, restore it to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or declare the breaker inoperable and apply ACTION 10.

The breaker shall not be bypassed while one of the diverse trip features is inoperable except for the time required for performing maintenance to restore the breaker to OPERABLE status.

MILLSTONE - UNIT 3 3/4 3-7 3

TABLE 3.3-2 3

REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES PS FUNCTIONAL UNIT RESPONSE TIME j

1.

Manual Reactor Trip N/A E

d5 0.5 second" 2.

Power Range, Neutron Flux u,

3 Power Range, Neutron Flux, N.A.

High Positive Rate 4.

Power Range, Neutron Flux,

[- 0.5 second' High Negative Rate 5.

Intermediate Range, Neutron Flux N.A 6.

Source Range, Neutron Flux N.A.

u,

's-7.

Overtemperature AT f 7 seconds' w

Ex>

8.

Overpowerab T di 7 seconds

9 Pressurizer Pressure--Low di 2 seconds 10.

Pressurizer Presssure--High f$,2 seconds 11.

Pressurizer Water Level--High N.A.

g S

$n

' Neutron detectors are exempt from response. time testing. Response time of the neutron flux signal portion of the me

?

channel shall be measured from detector output or input of first electronic component in channel.

N s

m

s TABLE 3.3-2 (Continued) 1

~~

f REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES S

oA FUNCTIONAL UNIT RESPONSE TIME e

12.

Reactor Coolant Flow--Low 5

a.

Single Loop (Above P-8)

< 1 secnnd b.

Two Loops (Above P-7 and below P-8)

< 1 second 13.

Steam Generator Water level--Low-Low

$ 2 seconds 14.

Low Shaft Speed-Reactor Coolant Pumps

$ 0.6 second**

15.

Turbine. Trip a.

Low Fluid Oil Pressure M.A.

b.

Turbine Stop Valve Closure N.A.

{

16.

Safety In.jection Input from ESF N.A.

"4 17.

Reactor Trip System Interlocks N.A.

18.

Reactor Trip Breakers N.A.

4 19.

Automatic Trip and Interlock logic M.A.

20.

Three Loop Operation Bypass Circuitry N.A.

S N

E;;g n

"Y

- Speed sensors are exempt from response time testing.

Response time of the speed signal portion of the

.P channel shall be measured from detector output or first electronic component in the channel.

,' oo

i TABLE ho3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS l

2 E'[;

TRIP g

ANALOG ACIITATING MODES FOR y

CIIANNEL DEVICE WiiICil CllANNEL CilANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE FUNCTIONAL UNIT CIIECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED e3 1.

Manual Reactor Trip N.A.

N.A.

R(14)

N.A.

1, 2, 3',

4',

S' 2.

Power Range, Nuetron Flux a.

liigh Setpoint S

D(2, 4),

Q(17)

N.A.

1, 2 M(3, 4),

Q(4, 6),

R(4, 5) b.

Low Setpoint S

R(4)

S/U(1)

N.A.

'N.A.

1***, 2 u,

?z 3

Power Range, Nuetron u,

1.

Flux, N.A.

R(4)

Q(17)

N.A.

1, 2 O

liigh Positive Rate 4.

Power Range, Neutron Flux N.A.

R(4)

High Negative Rate Q(17)

N.A.

1, 2

$g 5.

Intermediate Range, S

R(4, 5)

S/U(1)

N.A.

1***, 2 Ig 6.

Source Range, Neutron r+

Flux S

R(4, 5)

S/U(1),

N.A.

2,

3, 4, 5 ge Q(9, 17) 7.

Overtemperature Zh T S

R Q(17)

N.A N.A 1, 2 8.

Overpower ZL T S

R Q(17)

N.A.

1, 2 9

Pressurizer Pressure--

Low S

R Q(17, 18)

N.A.

1

De TABLE II.3-1 p2 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS M

4 s

TRIP c:

ANALOG ACUTATING MODES FOR 23 CHANNEL DEVICE WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE FUNCTIONAL UNIT CHECK CALIB RATION TEST TEST LOGIC TEST IS REQUIRED 10.

Pressurizer Pressure--

High S

R Q( 17, 18)

N.A.

1, 2' 11 Pressurizer Water Level--High 3

R Q(17)

N.A.

1 12.

Reactor Coolant Flow--

4 Low S

R Q(17)

N.A.

1 e

5

=

i M

N

TABLE 4.3-1 (Continued)

TABLE NOTATIONS When the Reactor Trip System breakers are closed and the Control Rod Drive System is capable of rod withdrawal.

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

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

(1)

If not performed in previous 31 days.

(2) Comparison of calorimetric to excore power indication above 15% of RATED THEF. MAL POWER.

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

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

'3)

Single point comparison of incore to excore AXIAL FLUX DIFFERENCE above 15% of RATED THERMAL POWER.

Recalibrate if the absolute difference is greater than or equal to 3%.

The provisions of Specification 4.0.4 are not applicable for entry into HUDE 2 or 1.

(4) Neutron detectors may be excluded from CHANNEL CALIBRATION.

(5) Detector plateau cur 0es shall be obtained, and evaluated and compared to manufacturer's date.

For the Intermediate Range and Power Range Neutron Flux channels the provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

(6) Incore - Excore Calibration, above 75% of RATED THERMAL POWER.

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

(7) Each train shall be tested at least every 62 d_ays on a STAGGERED TEST BASIS.

(8) (Not used)

(9) Quarterly surveillence in MODES 3*, 4*, and 5* shall also include verification that permissives P-6 and P-10 are in their reouired state for existing plant conditions by observation of the permissive annudciator window.

Quarterly surveillance shall include verification of the High Flux at Shutdown Alarm Setpoint of less than or equal to 5 times background.

i l

MILLSTONE - UNIT 3 3/4 3-13

TADLE 4.3-1-(Continued)-

TABLE NOTATIONS (Continued) l l

(10)

Setpoint verification is not applicable.

(11)'

The ThIP ACTUATING DEVICE OPERATIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip attachments of the

' Reactor Trip Breakere.

.(12)

(NOT-USED)

(13)

Reactor Coolant Pump Shaft Speed Sensor may be excluded from CHANNEL CALIBRATION.

r (14)

The TRIP ACTUATING DEVICE OPERATIONAL TEST shall independently verify the OPERABILITY of the undervoltage and stunt trip circuits for the Manual Reactor 1 rip Function.

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

(15)

Local manual shunt trip prior to placing breaker in service.

(16)

Automatic undervoltage trip.

(17)

Each channel shall be tested at least every 92 days en a STAGGERED TEST BASIS.

(18)

The surveillance frequency and/or MODES specified for these channels in Table 4 3-2 are more restrictive and, therefore, applicable.

t i

I l

MILLSTONE - UNIT 3 3/4 3-14 Amendment No.12 l

9 K

1 '

--w

- - ~ - - - -. - -

1

TABLE 3.3-4 (Continued) 1 2

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS r

G2 z

ni SENSOR TOTAL ERROR e

c:

FUNCTIONAL UNIT ALLOWANCE (TA)

Z 25

~

(S)

TRIP SETPOINT ALLOWABLE VALUE

]

3.

Containment Isolation (Continued)

2) Automatic Actuation Logic M.A.

N.A.

and Actuation Relays

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

b.

Phase "B" Isolation

}'

1) Manual Initiation N.A.

N.A.

i'

2) Automatic Actuation N.A.

N. A.

N.A.

E$

Logic and Actuation Relays

3) Containment Pressure--

3.3 1.01 1.75

< 8.0 psig

< 8.8 psig High-3 4.

Steam Line Isolation a.

Manual Initiation N.A.

N.A.

b.

Automatic Actuation Logic M.A.

M.A.

N.A.

and Actuation Relays c.

Containment Pressure--High-2 3.3 1.01 1.75

< 3.0 psig

< 3.8 psig d.

Steam Line Pressure--Low 17.7 15.31 2.2

> 658.6 psig*

> 644.9 psig*

e.

Steam Line Pressure -

5.0 0.5 0

< 100 psi /s**

< 122.7 psi /s**

Negative Rate--High

rp TABLE 3 3-4 (Continued)

G U

ENGINEERING SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIPS SETPOINTS M

e SENSOR e

TOTAL ERROR E5 FUNCTIONAL UNIT ALLOWANCE (TA)

_Z_

(S)

TRIP SETPOINT ALLOWABLE VALUE g

5.

Turbine Trip and Feedwater Isolation a.

Automatic Actuation Logic N.A

'!. A.

N.A.

N.A Actuation Relays 2.33 1.75 dE 82.0% or di 82.8% or narrow b.

Steam Generator Water 3.7 Level--High-High (P-14) narrow range range instrument instrument span.

span.

c.

Safety Injection See Item 1. above for all Sarety Injection Trip Setpoints and u,

Actuation Logic Allowable Valves.

U T ve Low Coincident u,

d.

a with Reactor Trip (P-4) oa o>

1) Four Loops Operating N.A.

N.A.

2: 5640F 2: 560.60F

2) Three Loops Operating N.A.

N.A.

2' 5640F di 560.60F 5

6.

Auxiliary Feedwater S

S a.

Manual Initiation N.A.

N.A.

5 se b.

Automatic Actuation Logic N.A.,

N.A.

?

and Actuation Relays c.

Steam Generator Water Level--Low-Low

1) Start Motor-Driven 20.5 18 98 1.75 JE 23.5% or 2 22.6% or narrow Pumps narrow range range instrument instrument span.

span.

c

TABLE 3.3-4 (Continued) 35 g

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS S

E SENSOR TOTAL ERROR g

FUNCTIONAL UNIT ALLOWANCE (TA) Z (S)

TRIP SETPOINT ALLOWABLE VALUE O

6.

Auxiliary Feedwater (Continued)

2) Start Turbme-(Iriven Ppmps 20.5 18.98 1.75

> 23.5X of

> 22.6% of narrow narrow range range instrument instrument span.

span.

d.

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

~

e.

Loss-of-Offsite Power N.A.

N.A.

> 2800V

> 2720V Start Motor-Driven Pumps O

f.

Containment Depressurization See Item 2. above for all CDA Trip Setpoints and Allowable Values.

Actuation (CDA) Start Motor-Driven Pumps 7.

Control Building Isola' tion a.

Manual Actuation N.A.

N.A.

b.

Manual Safety Injection N.A.

N.A.

M.A.

N.A.

Actuation c.

Automatic Actuation Logic N. A.

M. A.

N.A.

and Actuation Relays d.

Containment Pressure--High 1 3.3 1.01 1.75 5 3.0 psig

$ 3.8 psig e.

Control Building Inlet N.A.

N.A.

~< 1.5x10 5pc/cc < 1.5x10 5pc/cc Ventilation Radiation

~

f.

Outside Chierine High N.A.

N.A.

$ 5 ppm

~

$ 5 ppm

TABLE 3 3-4 (continued) x ENGINEERING SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIPS SETPOINTS U

SEP:SOR 5

TOTAL ERROR e

FUNCTIONAL UNIT ALLOWANCE (TA)

Z (S)

TRIP SETPOINT ALLOWABLE VALUE c

h 8.

Loss of Power a.

4 kV Bus Undervoltage N.A.

N.A.

E 2800 2 2720 volts (Loss of Voltage) volts with with a 62 a 6 2 second second time time delay.

delay.

b.

4 kV Bus Undervoltage N.A.

N.A.

t 3719 volts 2 3706 volts (Grid Degraded Voltage) with a 6 8 with a 6 8 second time second time delay with ESF.

delay with,ESF actuation or actuation or

& 300 second n 300 second u

D time delay time delay w

without ESF without ESF g

actuation.

9 Engineering Safety Features Actuation System Interlocks a.

Pressurizer Pressure, P-11 N.A.

N.A.

6 1985 psig f 1995 psig, a

=

b.

Low-Low Tug, P-12 N.A.

'N.A N.A.

2 SS30F 2_ 547.60F me c.

Reactor Trip, P-4 N.A.

N.A.

d.

Steam Generator Water Level, P-14 See Item S above for all Steam Generator Water Level Trip Setpoints and Allowable Values.

10.

Emergency Generator Load N.A.

N.A.

Sequencer w

TABLE 3.3-5 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONOS 4.

Steam Line Pressure--Low a.

Safety Injection (ECCS) 1 27(5)f37(4) 1)

Reactor Trip

<2 2)

Feedwater Isolation

< 6.8(3) 3)

Phase "A" Isolation 2(2)(6)jy2(1)(6) 4)

Auxiliary Feedwater 10 6

1 50(1) 5)

Service Water 6)

Start Diesel Generators

< 12 b.

Steam Line Isolation

[6.8(3) 5.

Containment Pressure--Rfgh-3 a.

Quench Spray

< 32(2)/42(1)

[2(2)(6)jg(1)(6) b.

Phase "B" Isolation c.

Motor-Driven Auxiliary Feedwater 5 60 P aps II) d.

Service Water ',

1 90 6.

Containment Pressure--High-2 a.

Steam Line Isolation 1 6.8(3) 7.

Steam Line Pressure - Negative Rate--High a.

Steas Line Isolation p

< 6.8(3) 8.

Staam Generator Water Level--High-High a.

Turbine Trip

< 2.5 1

[6.8(3) b.

Feedwater Isolation 9.

Staan Generator Water Level--Low-Low a.

Motor-Driven Auxiliary Teodwater Puses 5 60 b.

Turbine-Driven Auxiliary Feedwater P op 1 60 l

10.

Loss-of-Offsite Power a.

Motor-Driven Auxiliary Feedwater Pump i 60 i

i, MILLSTONE - UNIT 3 3/4 3-33 Amendment No. 3 APR 9 19 9

TABLE 3 3-5 (Continued)

ENOINEERED SAFETY FEATURES RESPONSE TIMES INITIATINO SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 11.

Loss of Power a.

4 kV Bus Undervoltage

- f 13 (Loss of Voltage) b.

4 kV Emergency Bus fb18(7)/310(8)

Undervoltage (Crid Degraded Voltage) 12.

Tave Low Coincident With Reactor Trip (P-4) a.

Feedwater Isolation f;12(3) 13.

Control Building Inlet Ventilation Radiation a.

Control Buildin's Isolation g@ 3 7 14 Outside Chlorine High b.

Control Building Isolation f

7 o

MILLSTONE - UNIT 3 3/4 3-34 Amendment No.19.12

REACTOR COOLANT SYSTEM ISOLATED LOOP LIMITING CONDITION FOR OPERATION 3.4.1.5 The RCS loop stop valves of-an isolated loop shall be shut and the power removed from the valve operators.

APPLICABILITY:

MODES 1, 2, 3 and 4.

ACTION:

With the requirements of the above specification not satisfied:

either shut the loop stop valves and remove power from the valve operators within one hour, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTOOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE ~ REQUIREMENTS 4.4.1.5 The RCS loop stop valves of an isolated loop shall be verified shut and power removed from the valve operators at least once per 31 days, t

l MILLSTONE - UNIT 3 3/4 4-7 i

)

c REACTOR COOLANT OYSTEM ISOLATED LOOP STARTUP LIMITING CONDITION FOR OPERATION 3 4.1.6 A reactor coolant loop shall re=ain isolated with power re=oved fr0=

the associated RCS loop stop valve operators until:

a.

The temperature at the cold les of the isolated loop is within 200F of the highest cold leg te:perature of the operating loops, b.

The boron concentration of the isolated loop is greater than.or equal l

to the boron concentration of the operating loops, or greater than 2300 pp: whichever is less c.

The isolated portion of the loop has been drained and is refilled, I

and d.

The reactor is saberitical by at least 1.6%Ak/k.

j APPLICABILITY: MODES 5 and 6.

l ACTION:

a.

With the requirements of the above specification not satisfied, do not open the isolated loop stop valves.

b.

The provisions of Specification 3.0.4 are not applicable.

SURVEItLANCE REQUIREMENTS 4.4.1.6.1 The isolated loop cold leg temperature shall be determined to be within 200F of the highest cold leg te=perature of the operating loops within 30 minutes prior to opening the cold leg stop valve..

4.4.1.6.2 The reactor shall be determined to be soberitical by at least 1.6%

Ak/K wit,hin 30 minutes prior to opening the cold leg stop valve.

l 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 l

loop shall be determined to:

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a.

Be drained and refilled, and b.

Have a boron concentration greater than or equal to the boron concentration of the operating loops, or greater than 2300 ppm whichever is less MILLSTONE - UNIT 3 3/44-8 Amendment No. 12 l

I 1

'5/4.5 EMERGENCY CORE C00 LINO SYSTEMS

. 3/4.5.1 ACCUMULATORS LIMITING CONDITION FOR OPERATION 3 5.1 Each Reactor Coolant System (ROS) accu =ulator shall be OPERABLE with:

a.

The isolation valve open and power removed, b.

A contained borated water volume of between 6618 and 6847 gallons, c.

A boron concentration of between 2200 and 2600 ppe, and d.

A nitrogen cover-pressure of between 636 and 694 psia.

APPLICABILITY: MODES 1, 2, and 3'.

ACTION:

a.

With one accumulator inoperable, except as a result of a closed isolation valve, restore the inoperable accumulator to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDSY 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 fo11owing.6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , b.

With one accu =u'lator inoperable due to the isolation valve being closed, either im:ediately open the isolation 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 accu =ulstor shall be de=enstrated 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 volume and nitrogen cover--

pressure in the tanks to be within the above limits, and 2)

Verifying that each accuculator isolation valve is open, 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 15 of tank volume by verifying the boron concentration of the accumulator solution; and l

' Pressurizer pressure above 1000 psig.

Amendment No.12 MILLSTONE - UNIT 3 3/4 5-1 i

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EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) c.

At least once per 31 days when the RCS pressure is above 1000 psig by verifying that power to the isolation valve operator is discon-nected by removal of the breaker from the circuit.

4.- 5.1. 2 Each accumulator water level and pressure channel shall be demon-Strated OPERABLE:

At lea',t once per 31 days by the performance of an ANALOG CHANNEL a.

OPERATIONAL TEST, and b.

At least once per 18 months by the performance of a CHANNEL CAllBRATION.

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MILLSTONE - UNIT 3 3/4 5-2

)

,v EMER0EN0Y CORE C0OLING SYSTEMS

~.

3/4.5.4 REFUELINO WATER STORACE TANK l

LIMITING 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 4

gallons, b.

A boron concentration between 2300 and 2600 ppm of boron.

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A minimu: solution te=perature of 400F,. and i

d.

A maximu= solution temperature of 500F.

APPLICAPILITf: HODES 1. 2. 3. and 4 i

ACTION 5

With the RWST inoperable, restore the tank to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STAND 3Y within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the Tollowing 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

i SURVEILLANCE REOUIREMENTS 4.5.4 The RWST shall be demonstrated OPERABLE:

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a.

At least once per 7 days by:

L 1)

Verifying the contained borated 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 MILLSTONE - UNIT 3 3/4 5-9 Amendment No. 12 d

1

CONTAINMENT SYSTEMS RECIRCULATION SPRAY SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.2 Two independent Recirculation Spray Systems shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTION:

With one Recirculation Spray Systim inoperable, restore the inoperable 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 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore the inoperable Recirculation Spray System to OPERABLE status within the next 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in COLD SHUTOOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.6.2.2 Each Recirculation Spray System shall be demonstrated OPERABLE:

a.

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

.or otherwise secured in position, is in its correct position; b.

By verifying, that on recirculation flow, each pump develops a differential pressure of greater than or equal to 130 psid when tested pursuant to Specification 4.0.5; c.

At least once per 18 months by verifying that on a CDA test signal, each recirculation spray pump starts automatically after a 660 2 20 second delay; d.

At least once per 18 months during shutdown', by verifying that each automatic valve in the flow path actuates to its correct position on a CDA test signal; and e.

At least once per 5 years by performing an air or smoke flow test through each spray header and verifying each spray nozzle is unobstructed, e

MILLSTONE - UNIT 3 3/4 6-13

CONTAINMENT SYSTEMS SPRAY ADDITIVE SYSTEM t

LIMITING CONDITION FOR OPERATION 3.6.2 3 The Spray Additive Syste: shall be OPERABLE with:

a.

A chemical addition tank containing a volume of between 1B000.and 19000 gallons of between 2.41 and 3 105 by weight NaOH solution, and-i b.

Two gravity feed paths each capable of adding NaOH solution from the chemical addition tank to each Containment Quench Spray subsystem j

pu=p suction.

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

With the Spray Additive Syste= inoperable, restore the syste= 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 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , 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 t

or be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REOUIREMENTS 4.6.2 3 The Spray Additive System shall be demonstrated OPERABLE:

At least once per 31 days by verifying that each valve (manual.

a.

power-operated, or auto:stic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position; b.

At least once per 6 months by:

1)

Verifying the contained solution volume in the. tank, and i

2)

Verifying the concentration of the NaOH solution by chemical analysis is within the above limits.

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

. autocatic valve in the flow path actuates to its correct position on a CDA test signal, i

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MILLSTONE - UNIT 3 3/4 6-14 Amendment No.12 i

1

. 3/b.9 REFUEL 8NG OPERATRONS i

3/h.9 1 BORON CONCENTRATION LIMITING CONDITION FOR OPERATION t

4 3 91.1 The boron concentration of all filled portions of the Reactor Coolant Syste= and the refueling canal shall be maintained uniform and sufficient to ensure that the more restrictive "of the following reactivity conditions is =et; either:

A X,77 of 0 95 or less, or a.

b.

A boron concentration of greater than or equal to 2300 pp=.

APPLICABILITY: MODE 6.'

ACTION:

With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and initiate and continue boration at greater than or equal to 33 sp:

of a solution containing greater than or equal to 6300 ppm boron or its equivalent until K is reduced to less than or equal to 0 95 or the boren concentrationisrIk(ocedtogreaterthanorequalto2300 ppm,whicheveris the more restrictive.

SURVEILLANOE REOUIREMENTS 4 9 1.1.1 The more restrictive of the above two reactivity conditions shall be determined prior to:

a.

Re=oving or unbolting the reactor vessel head, and b.

Withdrawal of any full-length control rod in exces's of 3 feet from 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 .

4.9 1.1 3 valve 3CHS-V305 shall be verified closed and secured in position by mechanical stops or by removal of air or electrical power at least once per 31 days.

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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.

MILLSTONE - UNIT 3 3/49-1 Amendment No.12

'9 REFUELING OPERATTONS i

BORON CONCENTRATION r

Limiting condition for Operation 3 9.1.2 The boron concentrction of the spent ruel Pool shall be maintained' uniform and sufficient to ensure that the boron concentration is greater than or equal to 800 ppm.

Applicability During ALL fuel asse=bly movements within the spent fuel pool, t

Action With the boron concentration less than 800 pps, suspend the move =ent of all fuel asse:blies within the spent fuel pool.

Surveillance Requirements j

4. 9.1.2 - Verify that the boron concentration is greater than or equal to 800 pp: prior to any movement of a fuel asse=bly into or within the spent f

fuel pool, and every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months thereafter during fuel move:ent.

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t MILLSTONE - UNIT 3 3/4 9-ta Arnendme;1t No.12 3

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REACT 8vITY CONTROL SYSTEMS BASES B0 RATION SYSTEMS (Continued)

MARGIN from expected operating conditions of 1.6% e k/k after xenon decay and cooldown to 2000F. The =axi=u= expected boration capability requirement occurs at EOL fro = full power equilibriu= xenon conditions and requires 21,020 gallons l

of 6300 pp= dore.ted water from the borie acid storage tanks or 1,166,000 gallons of 2300 pp: borated water fro = the refueling water storage tank (RWST). A mini =u= RWST volu=e of 1,166,000 gallons is specified to be consistent with ECCS requirement.

With the RCS te=perature below 2000F, one Boron 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 Syste= becomes inoperable.

The li=itation for a =axi=u: of one centrifugal charging pu=p to be OPER-ABLE and the Surveillance Require =ent to verify all charging pumps except the required OPERABLE pu=p to be inoperable below 3500F provides assurance that a

= ass addition pressure transient can be relieved by the operation of a single l

PORV.

The boron capability required below 2000F is sufficient to provide a SH'JTDOWN MARGIN of 1.6% 4 k/k after xenon decay and cooldown from 2000F to 1400F, Tnis condition requires either 4100 gallons of 6300 pp= borated water fro = the boric acid storage tanks or 250,000 gallons of 2300 pp: borated water fro = the RWST.

The contained water volume limits include allowance for wat,er not available because of discharge line location and other physical characteristics.

The li=its 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 LOOA. This pH band =inimites the evolution of iodine and =inimi:es the effect of chloride and caustic stress corrosion on

=echanical syste=s and compenents.

The =ini=u: RWST solution te=perature for MDDES 5 and 6 is based on analysis assu:ptions in addition to free:e protection considerations. The

=ini=u=/=axi=u: RWST solution te:peratures for MODES 1, 2, 3 and 4 are based on analysis assu:ptions.

The OPERABILITY of one Boron Injection Syste= during REFUELING ensures that this syste= is available for reactivity control while in MODE 6.

MILI.STOSE - UNIT 3 B 3 /4 1-3 Amendment No.12

r-4

~3/4.1 3 MOVA3LE CONTROL ASSEMBLIES i

The specifications of this section ensure that:

(1) acceptable power 7

distribation limits are maintained, (2) the.sinimum SHUTDOWN MARCIN is maintained, and (3) the potential effects of rod misaligr. ment on associated accident analyses are limited.

OPERABILITY of the control rod position-indicators is required to determine control-rod positions and thereby ensure

. compliance with-the control i

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l MILLSTONE - UNIT 3 B 3/4 13a Amendment No. 12 l

. POWER DISTRIBUTION LIMITS BASES 1

HEAT FLUX HOT CHANNEL FACTOR AND RCS FLOW RATE AND NUCLEAR EtiTHALPY RISE HOT CHANNEL FACTOR (Continued) c.

The control rod insertion limits of Specifications 3 1 3 5 and i

3.1 3.6 are maintained; and i

d.

The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, in maintained within the limits, i

F throughU.willbemaintainedwithinitslimitsprovideglonditionsa.

j above are maintained.

The relaxation of F*H as a function of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion limits.

The F1 as calculated in Specifications 3 2 31 and 3 2.3.2 are H

N used in the various accident analyses where F infivences parameters g

other than DNBR, e.g., peak clad te:perature, aNd thus is the maximum "as =easured" value allowed.

The difference betwegn the three and four-loop F,y esquations is due to Infour-lo$p,usedinthesafetyanalysesforthreg-loop a core restrictive F A operation.

operation, the allowable rteasured F ay calculated in Specificatibn 3.2.3.1 at 655 Raged Thermal Power is 6 1.65.

In three-loop operation, however, F" H is restricted to a censured value f 1.55 to be consistent with the safety analynes for three loop operation.

At zero power, both specifications allow the sa=e g

measured F4 H' Fuel rod bowing reduces the value of DNS ratio.

Credit is

available to offset this reduction in the generic margin. The generic margins, totaling 9.15 DNSR completely offset any rod bow penalties.

This cargin includes the following:

a.

Design limit DNBR of 1 30 vs 1.28, b.

Grid Spacing (Xs) of 0.046 vs. 0.059,

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c., Ther=al Diffusion Coefficient of 0.038 vs 0.059, d.

DNBR Multiplier of 0.86 vs. 0.88, and 4

e.

Pitch reduction.

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

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i MILLSTONE - UNIT 3 B 3/4 2-5 Amendment No. 12 i

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m-- -,

i POWER DISTRIBUTION LIMITS BASES HEAT FLUX HOT CHANNEL FACTOR AND RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued)

When an FQ measurement is taken, an allowance for both experimental error and manufacturing-tolerance uust be made.

An allowance of 55 is appropriate for a full-core map ta' ken with the Incore Detector Flux Mapping System, and a 35 allowanee is appropriate for manufacturing l

tolerance.

The Radial Peaking Factor, Fxy(2), it measured periodically to provideassurancethattheHotChannelFactor,Fy(Z)g)rcmainswithinits licit. The F limit for RATED THERMAL POWEB (FxpTP as provided in xy the Radial Peaking Factor Limit Report per Speciffcation 6 9 1.6 was deter ined from expected power control manuevers over the full range of burnap conditions in the core.

j N

I When RCS flow rate and F are measured, no additional allowances are necessary prior to conparMn with the limits of the Limiting Condition for Operation.

Measurement errors of 1.8% for four loog flow and 2.0% for three loop flow for RCS total flow rate and 45 for F have been allowed for in determination of the design DNBR value. g g i

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 cight not be detectd could bias the result fremthe precision heat balance in a nonconservative manner. Therefore, a penalty of 0.1% for undetected fouling of the feedwater venturi will be added if venturis are not verified clean every 18 months. Any fouling which cight bias the RCS flow rate ceasurement greater than 0.55 can be detected by monitoring and trending various plant performance para:eters.

IT detected, action shall be taken before perfor=ing subsequent precision heat balance measurements, i.e., either the effect j

of the fouling shall be quantified and com.ansated for in the RCS flow rate measure:ent or the venturi shall be cleaned to eliminate the fouling.

The 12-hour periodic surveillance of indicated RCS flow is sufficient to detect only flow degradation which could lead to operation outside the acceptable region of defined in Specification 3 2 3.1 and 3232.

3/4.2.4 CUADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power distribution satisfies the design values used in the power capability 1

analysis.

Radial power distribution measurements are made during STARTUP testing and periodically during power operation.

MILLSTONE - UNIT 3 B 3/4 2-6 Amendment No. 12 J

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- POWER D?STRIBUTTON LIM 2TS BASES QUADRANT POWER TILT RATIO (Continued)

The limit of 1.02, at which corrective action is required, provides DRB and linear heat generation rate protection with x-y plane power tilts. A limiting tilt of 1.025 r an be tolerated before the margin for uncertainty in Fg is depleted. A limit of 1.02 was selected to provide an allowance for the uncertainty associated with the indicated power tilt.

The 2-hour time allowance for operation with a tilt conditior, greater than 1.02 but less than 1.09 is provided to allow identificatLm and correction of a dropped or misaligned control rod.

In the event such action does not correct the tilt, the margin for uncertainty on Fg is reinstated by reducing the maximum allowed power by 31 for each percent of tilt in excess of 1.

For purposes of monitoring 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 00ADRANT POWER TILT RATIO. The incere detector monitoring is dont with a full incore flux map 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-11, H-3, H-13, L-5, L-11, N-8.

3/4 2.5 DNB PARAMETERS The limits on the DRB-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 of 130 throughout each analyzed transient. The indicated Tavg values are 591.20F (four loop operations) or 583.40F (three loops operating) and the indicated

{

pressurizer pressure value is 2226 psia (four loop or three loop i

operation). The calculated values of the DNB related parameters will be

)

an average of the indicated values for the operable channels.

The 12-hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.

Measurement uncertainties have been accounted for in determining the parameter limits.

MILLSTONE - UNIT 3 B 3/4 2-7 Amendment No.12 I

3/4 9 REFUELING OPERATIONS BASES 3 /4. 9.1 BORON CONCENTRATION The limitations on reactivity conditions during REFUELING ensure that:

(1) the reactor will re=ain suberitical during CORE ALTERATIONS, and (2) a unifor: boron concentration is maintained for reactivity control in the water volume having direct access to the reactor vessel. These limitations are consistent with the initial conditions assu=ed for the boron dilution incident in the safety analyses.

okk/k conservative allowance for uncertainties.Thevalueof095orlessforK((e includes a 15 Si=I

, the boren concentratien value of 2300 pp or greater includes a conservative uncertainty allowance of 50 pp= boron. The locking closed of the required valves during refueling operations precludes the possibility of uncontrolled boren dilution of the filled portion of the RCS. This action prevents flow to the ROS of unborated water by closing flow paths from sources of unborated water.

3/4.9.1.2 Boron Concentration in Spent Fuel Pool The limitations of this specification ensure that in the event of a fuel assembly handling accident involving eithvr a misplaced or dropped fuel assembly, the K rr of the spent fuel storage racks will re=ain less than or e

equal to.95.

3/4.9 2 INSTRUMENTATION The OPERABILITY of the Source Range Neutron Flux Monitors ensures that redundant monitoring capability is available to detect changes in the l

reactivity condition of the core.

3/4.9.3 DECAY TIME The minimum requirement for reactor suberiticality prior to movement of I

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 ti=e is consistent with the assumptions u' sed in the safety analyses.

l 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS

{

The requirements on containment building penetration closure and l

I OPERABILITY ensure that a release of radioactive material within containment 4

will bS restricted from leakage to the environment.

The OPERABILITY and l

closure restrictions are sufficient to restrict radioactive material release fro: a fuel element rupture based upon the lack of containment pressurization potential while in the REFUELING 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 facility status or core reactivity conditions during CORE ALTERATIONS.

MILLSTONE - UNIT 3 5 3/4 9-1 Amendment No. 12 l

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l REFUELING OPERATIONS BASES l

3/4.9.6 REFUELING MACHINE I

The OPERABILITY requirements for the refueling machine ensure that:

(1) refueling machines will be used for movement of drive rods and fuel assem-blies, (2) each crane has sufficient load capacity to lift a drive rod or fuel assembly, and (3) the core internals and reactor vessel are protected from excessive lifting force in the event they are inadvertently engaged during l

lifting operations.

3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE AREAS The restriction on movement of loads in excess of the nominal weight of a i

fuel and control rod asse cly and associated handling tool over other fuel assemblies in the storage pool ensures that in the event this load is dropped:

(1) the activity release will be limited to tnat contained in a single fuel assembly, and (2) any possible distortion of fuel in the storage racks will not result in a critical array.

This assumption is consistent with the activity release assumed in the safety analyses.

3/4.9.8 RE510 VAL HEAT REMOVAL AND COOLANT CIRCULATION l

The requirement that at least one residual heat removal (RHR) loop be in i

operation ensures that:

(1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor vessel below 140*F as required l

during the REFUELING PODE, and (2) sufficient coolant circulation is maintained through the core to minimize the effect of a boron dilution incident and prevent boron stratification.

[

The requirement to have two RHR loops OPERABLE when there is less than 23 feet of water above the reactor vessel flange ensures that a single failure of the operating RHR loop will not result in a complete loss of residual heat removal capability.

With the reactor vessel head removed and at least 23 feet of water above the reactor pressure vessel flange, a large heat sink is avail-able for core cooling.

Thus, in the event of a failure of the operating i

RHR loop, adequate time is provided to initiate emergency procedures to cool the core. **

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3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM The OPERABILITY of this system ensures that the containment vent and purge penetrations will be automatically isolated upon detection of high radiation levels within the containment.

The OPERABILITY of this system is required to restrict the release of radioactive material from the containment atmosphere to the environment.

MILLSTONE - UNIT 3 8 3/4 9-2

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t DESIGN FEATURES

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53 REa: TOR CORE 4

FUEL ASSEMBLIES i'

531 The core shall contain 193 fuel assemblies with each fuel asse=bly containing 264 fuel rods clad with Zircaloy-4 Each fuel rod shall have a nominal active fuel length of 144 inches.

The initial core loading shall have i

a =aximum nominal enrichment of 3.4 weight percent U-235.

Reload fuel shall l

be similar in physical design to the initial core loading and shall have a maximu nominal enrichment of 3 8 weight percent U-235.

a CONTROL RCD ASSEM3 TIES i

532 The core shall contain 61 full-length control rod asse:blies. The full-length control rod asse:blies shall contain a nominal 142 inches of absorber caterial. The nominal values of absorber material shall be 95 35 1

harnium and 4 55 natural zirconium.

All control rods shall be clad with stainless steel.

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4 5.4 REACTOR COOLANT SYSTEM i

i CESIGN PRESSURE AND TEMPERATURE 5.4.1 Tne Reactor coorant Syste= is designed and shall be maintained:

a.

In accordance with the Code require =ents specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the 3

j applicable Surveillance Requirements,

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j b.

For a pressure of 2500 psia, and i

c.

For a temperature or 6500F, except for the pressurizer which is 6B00F.

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VOLUME 5.4.2 The total water and steam volume of the Reactor coolant System is I

12,240 cubic feet at a nominal T,y, of 5870F.

i 5.5 METEOROLOGICAL TOWER LOCATION 1

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5 5.1 The meteorolegical tower shall be located as shown on Figure 5.1-3 i

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MILLSTONE - UNIT 3 5-5 Amendment No.12

a DESIGN FEATURES 5.6 FUEL STORAGE CRITICALITY

5. 6.1.1 The spent fuel storage racks are designeo and shall be maintained with:

A k,gg equivalent to less than or equal to 0.95 when flooded with 4.

unborated water, which includes a conservative allowance of 2.6%

ok/k for uncertainties as described in Section 4.3 of the FSAR, and b.

A nominal 10.35-inch center-to-center distance between fuel assemblies placed in the storage racks.

5.6.1.2 The k,ff for new fuel or the first core loading stored dry in the spent fuel storage racks shall not exceed 0.98 when aqueous foam moderation is assumed.

ORAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 45 feet.

CAPACITY 5.6.3 The spent fuel storag: pool is designed and shall be maintained with a storage capacity limited to no more than 756 PWR fuel assemblies.

5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT 5.7.1 The components identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7-1.

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i MILLSTONE - UNIT 3 5-6 l

ML20149C916

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