ML082250418

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Stretch Power Uprate, Tec Specs to Amd 242
ML082250418
Person / Time
Site: Millstone Dominion icon.png
Issue date: 08/12/2008
From:
NRC/NRR/ADRO/DORL/LPLI-2
To:
Lamb John G./NRR/DORL, 415-3100
Shared Package
ml082180137 List:
References
TAC MD6070
Download: ML082250418 (23)
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Text

B, Subject to the conditions and requirements incorporated herein, the Commission hereby licenses; (1)

DNC, pursuant to Section 103 of the Act and 10 CFR Part 50, to possess, use and operate the facility at the designated location in New London County, Connecticut in accordance with the procedures and limitations set forth in this license; Central Vermont Public Service Corporation and Massachusetts Municipal Wholesale Electric Company, pursuant to the Act and 10 CFR Part 50, to possess the facility at the designated location in New London County, Connecticut in accordance with the procedures and limitations set forth in this renewed operating license; (2)

DNC, pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (3)

DNC, pursuant to the Act and 10 CFR Parts 30, 40, and 70 to receive, possess, and use at any time any byproduct, source, and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation. monitoring equipment calibration, and as fission detectors in amounts as required; (4)

DNC, pursuant to the Act and 10 CFR Parts 30, 40, and 70 to receive, possess, and use in amounts as required any byproduct, source, or special nuclear material without restriction to chemical or physical form for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5)

DNC, pursuant to the Act and 10 CFR Parts 30, 40, and 70 to possess, but not separate, such byproducts and special nuclear materials as may be produced by the operations of the facility.

C.

This renewed operating license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter 1 and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1)

Maximum Power Level DNC is authorized to operate the facility at reactor core power levels not in excess of 3650 megawatts thermal (100 percent rated power) in accordance with the conditions specified herein.

Renewed License No. NPF-49 Amendment No. 242 (2)

Technical Specifications The Technical Specifications contained in Appendix A, revised through Amendmen't No. 242 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto are hereby incorporated into the renewed license. DNC shall 0perate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3)

DNC shall not take any action that would cause Dominion Resources, inc.

(DRI) or its parent companies to void, cancel,. or diminish DNC's commitment to have sufficient funds available to fund an extended plant shutdown as represented in the application for approval of the transfer of the licenses for MPS Unit No. 3.

(4)

Immediately after the transfer of interests in MPS Unit No. 3 to DNC, the amount in the decommissioning trust fund for MPS Unit No. 3 must, with respect to the interest in MPS Unit No. 3, that DNC would then hold, be at a level no less than the.formula amount under 10 CFR 50.75.

(5)

The decommissioning trust agreement for MPS Unit no. 3 at the time the transfer of the unit to DNC is effected and thereafter is subject to the following:

(a)

The decommissioning trust agreement must be in a form acceptable to the NRC.

(b)

With respect to the decommissioning trust fund, investments in the securities or other obligations of Dominion Resources, Inc. or its affiliates or subsidiaries, successors, or assigns are prohibited.

Except for investments tied to market indexes or other non-nuclear-sector mutual funds, investments in any entity owning one or more nuclear power plants are prohibited.

(c)

The decommissiong trust agreement for MPS Unit No. 3 must provide that no disbursement or payments from the trust, other than for ordinary administrative expenses, shall be made by the trustee until the trustee has first given the Director of the Office of Nuclear Reactor Regulation 30 days prior written notice of payment. The decommissioning trust agreement shall further contain a provision that no disbursements or payments from the trust shall be made if the trustee receives prior written notice of objection from the NRC.

(d)

The decommissioning trust agreements must provide that the agreement can not be amended in any material respect without 30 days prior written notification to the Director of the Office of Nuclear Reactor Regulation.

Renewed License No. NPF-49 Amendment No. 242

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION 3/4.6.2 3/4.6.3 3./4.6.4 3/4.6.5 3/4.6.6 3/4.7 PLANT 3/4.7.1 TABLE 3.7-1 TABLE 3.7-2 PAGE Air Temperature...............

3/4 6-9 A ir T m p e ra u re............,...

........ 3 4 -

Containment Structural Integrity..............................

... 3/4 6-10 C ontainm ent Ventilation System..............................................................

3/4 6-11 DEPRESSURIZATION AND COOLING SYSTEMS Containm ent Quench Spray System.........................................................

3/4 6-12 R ecirculation Spray System.......................................................................

3/4 6-13 CONTAINMENT ISOLATION VALVES..................................................

3/4 6-15 DELETED SUBATMOSPHERIC PRESSURE CONTROL SYSTEM Steam Jet Air Ejector 3/46-18 SECONDARY CONTAINMENT Supplementary Leak Collection and Release System................................

3/4 6-19 Secondary C ontainm ent..........................................................................

3/4 6-22 Secondary Containment Structural Integrity..........

3/4 6-23 SYSTEMS TURBINE CYCLE Safet, Valves 3/4 7-1 OPERABLE MSSVs VERSUS MAXIMUM ALLOWABLE POWER.... 3/4 7-2 D E L E T E D...................................................................................................

3/4 7-2 I

MILLSTONE - UNIT 3 ix Amendment No. 59g, 63, 3-7g,99, -l0, 4--5, 1--26, 2-1-,, 2_24, 242

DEFINITION S PHYSICS TESTS 1.21 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrurnentation: (1) described in Chapter 14.0 of the FSAR, (2) authorized under the provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission.

1.22. DELETED PURGE - PURGING 1.23 PURGE or PURGING shall be any controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

QUADRANT 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 calibrated 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 detector inoperable, the remaining three detectors shall be used for computing the average.

RATED THERMAL POWER 1.27 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3650 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. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC.

REPORTABLE EVENT 1.29 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 of 10 CFR Part 50.

SHUTDOWN MARGIN 1.30 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming all ful-l-length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn.

MILLSTONE - UNIT 3 1-5 Amendment No. 69, 4-8+, 4--89, 2--4, 242

120 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, Reactor Coolant System highest loop average temperature, and pressurizer pressure shall not exceed the limits specified in the CORE OPERATING LIMITS REPORT; and the following Safety Limits shall not be: exceeded:

2.1:1.1 The departure from nucleate boiling ratio (DNBR) shall be maintained greater than or equal to 1.14 for the WRB-2M DNB correlations.

2.1.1.2 The peak fuel centerline temperature shall be maintained less than 5080'F, decreasing by 58°F per 10,000 MWD/MTU of bunmup.

APPLICABILITY:

MODES I and 2.

ACTION:

Whenever the Reactor Core Safety Limit is violated, restore compliance and be in HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2750 psia.

APPLICABILITY:

MODES 1, 2, 3, 4, and 5.

ACTION:

MODES 1 and 2:

Whenever the Reactor Coolant System pressure has exceeded 2750 psia be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

MODES 3, 4 and 5:

Whenever the Reactor Coolant System pressure has exceeded 2750 psia, reduce the Reactor Coolant System pressure to within its I imit within 5 minutes.

MILLSTONE - UNIT 3 2-1 Amendment No. 4-7-, 2-44, -436, 2 4 2

H' 0z:

TABLE 2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINT*

NOMINAL TRIP SETPOINT FUNCTIONAL uNrr 8.

9.

10.

11.

Overpower AT Pressurizer Pressure-Low Pressurizer Pressure-High Pressurizer Water Level-High

12. Reactor Coolant Flow-Low
13. Steam Generator Water Level Low-Low
14. General Warning Alarm
15. Low Shaft Speed - Reactor Coolant Pumps See Note 3 1900 psia 2385 psia 89% of instrument span 90% of nominal loop flow 18.1% of narrow range instrument span N.A.

92.4% of rated speed S,

ALLOWA13LE VALUE See Note 4

_ 1897.6 psia

< 2387.4 psia 89.3*%,, of instrument span

_ 89.8% of nominal loop flow

_ 17.8% of narrow range instrument span N.A.

_ 92.2% of rated speed rtD CD 0.

Zz TABLE 2.24 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOMINAL m FUNCTIONAL UNIT

16.

Turbine Trip

a.

Low Fluid Oil Pressure

b.

Turbine Stop Valve Closure

17.

Safety Injection Input from ESF 1 8.

Reactor Trip System Interlocks

a.

Intermediate Range Neutron Flux, P-6

b.

Low Power Reactor Trips Block, P-7

1) Power Range Neutron Flux, P-10 input (Note 5)
12)

Turbine Impulse Chamber Pressure, P-13 input 2

c.

Power Range Neutron Flux, P-8 S** RTP = RATED THERMAL POWER LJ NJ TRIP SETPOINT 500 psig 1% open N.A.

1 x 10-° amp 11% of RTP**

10% RTP** Turbine Impulse Pressure Equivalent 50.0% of RTP**

ALLOWABLE VALUE

_ 450 psig

Ž 1% open N.A.

_9.0 x 10- 1 amp

  • 11.6% of RTP**

_ 10.6% RTP**

Turbine Impulse Pressure Equivalent

< 50.6% of RTP**

TABLE 2.2-1 (Continued) o TABLE NOTATIONS z

SNOTE 3:

OVERPOWER AT AT

(+r IS)

T) (1 +- 2s < K4 -K 6 (T - T")

Where:

AT is measured Reactor Coolant System AT, OF; AT0 is loop specific indicated AT at RATED THERMAL POWER, OF; (1 + Ts).

(1 + -- s) is the function generated by the lead-lag compensator on measured AT; TI and -c,, are the time constants utilized in the lead-lag compensator for AT, -r _ [*]sec, T _< [*] sec; K4 * [*];

T is measured average Reactor Coolant System temperature, OF; T" is loop specific indicated Tavg at RATED THERMAL POWER, < [*]°F; 6 Ž> [*]/OF when T > T" and K6 _ [*]/OF when T < T";

CD s is the Laplace transform operator, sec o(The values denoted with are specified in the COLR.)

t'j

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

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

a.

RCS total.flow rate > 363,00 gpm and greater than or equal to the limit. specified in the CORE OPERATING LIMITS REPORT (COLR), and N

RTP

b.

FAH < FAH

[1.0 + PFAH(1.0- P)]

Where:

THERMAL POWER RATED THERMAL POWER' N

N

2)

FAH Measured values of FAH obtained by using the movable incore detectors to obtain a power distribution map. The measured value N

of FAH should be used since Specification 3.2.3.lb. takes into consideration a measurement uncertainty of 4% for incore measurement, RTP N

3)

FAH

= The FAH limit at RATED THERMAL POWER in the COLR, N

4)

PFAH= The power factor multiplier for FAH provided in the COLR, and

5)

The measured value of RCS total flow rate shall be used since uncertainties of 2.4% for flow measurement have been included inSpecification 3.2.3. 1a.

APPLICABILITY:

MODE I.

ACTION:

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

a.

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

1.

Restore the RCS total flow rate to within the limits specified above and in

'N the COLR and FAH to within the above limit, or

.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 <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

MILLSTONE - UNIT 3 3/4 2-19 Amendment No., 2O, 60, 444, 2--7-,

2-6, 242

POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION ACTION: (Continued)

b.

Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of initially being outside the above limits, verify through incore flux mapping and.RCS total flow rate that the RCS total flow rate is N

restored to within the limits specified above and in the COLR and FAH is restored to within the above limit, or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

c.

Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced THERMAL POWER limit required by ACTION a.2. and/or b., above; subsequent N

POWER OPERATION may proceed provided that FAH and indicated RCS total flow rate are 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:

I.

A nominal 50% of RATED THERMAL POWER,

2.

A nominal 75% of RATED THERMAL POWER, and

3.

Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 applicable.

N 4.2.3.1.2 FAH shall be determined to be within the acceptable range:

a.

Prior to operation above 75% of RATED THERMAL POWER after each fuel loading, and At leastonce per 31 I Effective Full Power Days.

4.2.3.1.3 The RCS total flow rate shall be determined to be within the acceptable range by:

a.

Verifying by precision heat balance that the RCS total flow rate is

> 363,200 gpm and greater than or equal to the limit specified in the COLR within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reaching 90% of RATED THERMAL POWER after each fuel loading, and MILLSTONE - UNIT 3 3/4 2-20 Amendment No. 60, -79, N00, 6, 2 4 2

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued)

b.

Verifying that the RCS total flow rate is Ž 363,200 gpm and greater than or equal to the limit specified in the COLR at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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

4.2.3.1.5 DELETED.

4.2.3.1.6 DELETED.

MILLSTONE - UNIT 3 3/4 2-21 Amendment No. 2-7, 60, 242

CIO

-A 0

TABLE 3.3-3 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TOTAL NO.

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

9.

Engineering Safety Features Actuation System Interlocks

a.

Pressurizer Pressure, P-11

b.

Low-Low Tavg, P-12

c.

Reactor Trip, P-4

10.

Emergency Generator

-Load Sequencer

11.

Cold Leg Injection Pernissive, P-19 3

4 2

2 4

2 2

2 2

3 2

2 3

1,2,3 1 2, 3 1,2,3 1,2,3,4 1,2 3 21 23 15 20 1

2 z

0D p

TABLE 3.3-3 (Continued)

TABLE NOTATIONS

'The Steamline Isolation Logic and Safety Injection.Logic for this trip function may be blocked in this MODE below the P-il (Pressurizer Pressure interlock) Setpoint.

MODES 1, 2, 3, and 4.

During fuel movement within containment or the spent fuel pool.

Trip function automatically blocked above P-Il and may be blocked below P-II when Safety Injection on low steam line pressure is not blocked.

ACTION STATEMENTS ACTION 14-ACTION 15 -

ACTION 16-ACTION 17 -

ACTION 18 -

ACTION 19 -

With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and, in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />; however, one channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing per Specification 4.3.2.1, provided the other channel is OPERABLE.

With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />; however, one channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing per Specification 4.3.2., provided the other channel is OPERABLE.

With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed until performance of the next required ANALOG CHANNEL OPERATIONAL TEST provided the inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed provided the inoperable channel is placed in the bypassed condition and the Minimum Channels OPERABLE requirement is met. One additional channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing per Specification 4.3.2.1.

With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 7 days.

After 7 day-.s, or if no channels are OPERABLE, immediately suspend fuel movement, if applicable, and be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

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 <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

MILLSTONE - UNIT 3 3/4 3-24 Amendment No. -54, 7-,

9, 8--29, 203, 2-, q-4, 242

C/

TABLE 3.3-4 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS NOMINAL ALLOWABLE*

JNCTIONAL UNIT TRIP SETPOINT VALUE Fi

8.

Loss of Power

a. 4 kV Bus Undervoltage (Loss of Voltage)

WJ

-41

b. 4 kV Bus Undervoltage (Grid Degraded Voltage)
9.

Engineered Safety Features Actuation System Interlocks

a. Pressurizer Pressure, P-11
b. Low-Low Tavg, P-12
c. Reactor Trip, P-4
10. Emergency Generator Load Sequencer
11. Cold Leg Injection Permissive, P-19 2800 volts with a

< 2 second time delay.

3730 volts with a

< 8 second time delay with ESF.

actuation or < 300 second time delay without ESF actuation.

1999.7 psia 553 0F N.A.

N.A..

1900 psia

> 2720 volts with a <_ 2 second time delay.

_ 3706 volts with a < 8 second time delay with ESF actuation or < 300 second time delay without ESF actuation.

< 2002.1 psia

> 552.60F N.A.

N.A.

> 1897.6 psia z

TABLE 4.3-2 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS Cl) 0z ANALOG CHANNEL CHANNEL CHANNEL OPERATIONAL CHECK CALIBRATION TEST TRIP ACTUATING DEVICE OPERATIONAL TEST MODES MASTER SLAVE FOR WHICH RELAY RELAY SURVEILLANCE TEST TEST IS REOUIRED ACTUATION LOGIC TEST FUNCTIONAL UNIT

7. Control Building Isolation (Continued)
e.

Control Building Inlet Ventilation Radiation S

R Q

N.A.

N.A.

N.A.

N.A.

8. Loss of Power CD z
a.

4 kV Bus Undervoltage (Loss of Voltage)

b. 4 kV Bus Undervoltage (Grid Degraded Voltage)
9. Engineered Safety Features Actuation System Interlocks
a. Pressurizer Pressure, P-11
b. Low-Low Tavg, P-12
c.

Reactor Trip, P-4

10. Emergency Generator Load Sequencer
11. Cold Leg Injection Permissive, P-19 N.A.

N.A.

R R

N.A.

N.A.

M(3)

M(3)

N.A.

N.A.

N.A.

N.A.

1, 2, 3, 4 N.A.

N.A. _

, 2, 3, 4 N.A.

N.A.

N.A.

N.A.

S R

R Q

Q N.A.

N.A.

R N.A.

N.A.

N.A.

N.A.

N.A.

N.A.

N.A.

1,2,31 N.A.

N.A.

1,2,.3 N.A.

N.A.

1,2,3 N.A.

N.A.

R N.A.

N.A.

Q Q(1,2)

N.A.

N.A.

1,2,3,4 N.A.

N.A.

N.A.

1,2,3

PRESSURIZER LEVEL CONTROL 80 70 60 2'

50 40 30 20 10 0

550 555 560 565 570 575 580 585 590 T(AVG), -F FIGURE 3.4-5 MILLSTONE - UNIT 3 3/4 4-11 a Amendment No. 4-60, 242

3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE SAFETY VALVES LIMITING CONDITION. FOR OPERATION 3.7.1.1 All main steam line Code safety valves (MSSVs) shall be OPERABLE with lift settings as specified in Table 3.7-3.

APPLICABILITY:

MODES 1, 2, and 3.

ACTION:

NOTE -----------------

Separate Condition entry is allowed for each MSSV.

a.

With one or more steam generators (SGs) with one MSSV inoperable, and the Moderator Temperature Coefficient (MTC) zero or negative at all power levels, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> reduce THERMAL POWER to less than or equal to 60.1 % RATED THERMAL POWER (RTP); otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b.

With one or more SGs with two or more MSSVs inoperable, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> reduce THERMAL POWER to less than or equal to the maximum allowable % RTP specified in Table 3.7-1 for the number of OPERABLE MSSVs, and reduce the Power Range Neutron Flux High setpoint to less than or equal to the maximum allowable % RTP specified in Table 3.7-1 for number of OPERABLE MSSVs within the next32 hours*; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

c.

With one or more SGs with one MSSV inoperable and the MTC positive at any power level, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> reduce THERMAL POWER to less than or equal to the maximum allowable % RTP specified in Table 3.7-1 for the number of OPERABLE MSSVs and reduce the Power Range Neutron Flux High setpoint to less than or equal to the maximum allowable % RTP specified in Table 3.7-1 for number of OPERABLE MSSVs within the next 32 hour3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br />s*; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Applicable only in MODE 1.

MILLSTONE - UNIT 3 3/4 7-1 Amendment No. --7, 2-1-4, 2 4 2

3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE SAFETY VALVES LIMITING CONDITION FOR OPERATION ACTION:

(Continued)

d.

With one or more SGs with four or more MSSVs inoperable, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.7.1.1 No additional Surveillance Requirements other than those required by Specification 4.0.5. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3.

MILLSTONE - UNIT 3 3/4 7-1a Amendment No. 24f

TABLE 3.7-1 OPERABLE MSSVs VERSUS MAXIMUM ALLOWABLE POWER NUMBER OF OPERABLE MSSVs

-. PER STEAM GENERATOR MAXIMUM ALLOWABLE POWER, (PERCENT OF RATED THERMAL POWER) 60.1 42.8 4

2 25.5 TABLE 3.7-2 DELETED MILLSTONE - UNIT 3 3/4.7-2 Amendment No. 4-042, 2+7, 2 4 2

PLANT SYSTEMS 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.7.7 7 Two independent Control Room Emergency Air Filtration Systems shall be OPERABLE.#.

APPLICABILITY:

MODES 1, 2, 3, and 4.

During fuel movement within containment or the spent fuel pool.

ACTION:

MODES 1, 2, 3 and 4:

a.

With one Control Room Emergency Air Filtration System inoperable, restore the inoperable system to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b.

With both Control Room Emergency Air Filtration Systems inoperable, except as specified in ACTION c., immediately, suspend the movement of fuel within the spent fuel pool.* Restore at least one inoperable system to OPERABLE status within 1' hour or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN vuithin the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

c.

With both Control Room Emergency Air Filtration Systems inoperable due to an inoperable Control Room boundary, immediately suspend the movement of fuel within the spent fuel pool and restore the Control Room boundary to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

During fuel movement within containment or the spent fuel pool:

d.

With one Control Room Emergency Air Filtration System inoperable, restore the inoperable system to OPERABLE status within 7 days. After 7 days, either initiate and maintain operation of the remaining OPERABLE Control Room Emergency Air Filtration System in the recirculation mode of operation, or immediately suspend the movement of fuel.

e.

With both Control Room Emergency Air Filtration Systems inoperable, or with the OPERABLE Control Room Emergency Air Filtration System required to be in the recirculation mode by ACTION d. not capable of being powered by an OPERABLE emergency power source, immediately suspend the movement of fuel.

The Control Room boundary may be opened intermittently under administrative control.

MILLSTONE - UNIT 3 3/4 7-15 Amendment No. 2, 4-8+, 203, 2-4-1, 2 4 2

TABLE 3.7-6 (Continued)

AREA TEMPERATURE MONITORING AREA TEMPERATURE LIMIT ('F)

7.

FUEL BUILDING FB-02, Fuel Pool Pump Cubicles, El 241 6"

< 119 FB-03, General Area, El 52' 4"

< 108

8.

FUEL OIL VAULT FV-01, Diesel Fuel Oil Vault

< 95

9.

HYDROGEN RECOMBINER BUILDING HR-01, Recombiner Skid Area, El 24' 6"

< 125 HR-02, Controls Area, El 24' 6"

< 110 HR-03, Sampling Area, El 24' 6"

<110 HR-04, HVAC Area, El 37' 6"

110
10.

MAIN STEAM VALVE BUILDING MS-01, Areas above El. 58' 0"

  • 140 MS-02, Areas below El. 58' 0"
  • 140
11.

DELETED

12.

TUNNEL TN-02, Pipe Tunnel-Auxiliary, Fuel and

  • 112 ESF Building
13.

YARD YD-01, Yard

  • 115 MILLSTONE - UNIT 3 3/4 7-35 Amendment No. 92, 4-00, 4-1K, 2242

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued).

2)

Pre-planned operating procedures and backup instrumentation to be used if one or more monitoring instruments become inoperable, and"

3)

Administrative procedures for returning inoperable instruments to OPERABLE status as soon as practicable.

f.

Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions.*. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163,."Performance-Based Containment Leak-Test Program,"dated September 1995, as modified by the following exception to NEI 94701, Rev. 0, "Industry Performance Based Option of 10 CFR Part 50 Appendix J": The first Type A test performed after. the January 6, 1998 TypeA test shall be performed no later than January 6, 2013.

The peak calculated containment internal pressure for the design basis loss of coolant accident, Pal is 41.4 psig..

The maximum allowable containment leakage rate La, at Pa, shall be 0.3 percent by weight of the containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Leakage rate acceptance criteria are:

1)

Containment overall leakage rate acceptance criterion is

  • 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the combined Type B and Type C tests, and < 0.06 La for all penetrations that are Secondary Containment bypass leakage paths, and < 0.75 La for Type A tests;
2)

Air lock testing acceptance criteria are:

a.

Overall air lock leakage rate is

  • 0.05 La when tested at > Pa
b.

For each door, seal leakage rate is < 0.01 La when pressurized to >_ Pa.

The provisions of Specification 4.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program.

The provisions of Specification 4.0.3 are applicable to the Containment Leakage Rate Testing Program.

An exemption to Appendix J, Option A, paragraph III.D.2(b)(ii), of 10 CFR Part 50, as approved by the NRC on December 6, 1985.

MILLSTONE - UNIT 3 6-17 Amendment No. 69, 4-86, 2-3-2, 23-9, 2 4 2

ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Cont.)

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.EVALUATION METHODOLOGY," (W Proprietary). (Methodology for Specifications

2. 1. 1. 1--Departure from Nucleate Boiling Ratio, 2:1.1.2--Peak Fuel Centerline Temperature, 3.1.1.3--Moderator Temperature Coefficient, 3.1.3.5--Shutdown Bank Insertion Limit, 3.1.3.6--Control Bank 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, 3.1.1.1.1, 3.1.1.1.2, 3.1.1.2-- SHUTDOWN MARGIN, 3.9.1.1 -- Boron Concentration.)
2.

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.

3.

NUREG-800, Standard Review Plan, U.S. Nuclear Regulatory Commission, Section 4.3, Nuclear Design, July 1981 Branch Technical Position.CPB 4.3-1, Westinghouse Constant Axial Offset Control (CAOC), Revision 2, July 1981.

4.

WCAP-10216-P-A-R1A, "RELAXATION OF CONSTANT AXIAL OFFSET CONTROL FQ SURVEILLANCE TECHNICAL SPECIFICATION,"

(WProprietary). (Methodology for Specifications 3.2.1--AXIAL FLUX DIFFERENCE [Relaxed Axial Offset Control] and 3.2.2--Heat Flux Hot Channel Factor [W(' ) surveillance requirements for EQ Methodology].)

5.

WCAP-12945-P-A, "CODE QUALIFICATION DOCUMENT FOR BEST.

ESTIMATE LOCA ANALYSIS,"

(W Proprietary). (Methodology for Specification 3.2.2--Heat Flux Hot Channel Factor.)

6.

WCAP-16009-P-A, "REALISTIC LARGE-BREAK LOCA EVALUATION METHODOLOGY USING THE AUTOMATED STATISTICAL TREATMENT OF UNCERTAINTY METHOD (ASTRUM)," (W Proprietary). (Methodology for Specification 3.2.2--Heat Flux Hot Channel Factor.)

7.

WCAP-11946, "Safety Evaluation Supporting a More Negative EOL Moderator Temperature Coefficient Technical Specification for the Millstone Nuclear Power Station Unit 3," (W Proprietary).

8.

WCAP-10054-P-A, "WESTINGHOUSE SMALL BREAK ECCS EVALUATION MODEL 17 USING THE NOTRUMP CODE," (W Proprietary). (Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)

9.

WCAP-10079-P-A, "NOTRUMP - A NODAL TRANSIENT SMALL BREAK AND GENERAL NETWORK CODE," (W Proprietary). (Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)

10. WCAP-12610, "VANTAGE+ Fuel Assembly Report," (_W Proprietary).

(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)

MILLSTONE - UNIT 3 6-20 Amendment No. 2-4, -3, 60, 69, 8+-,

2, -q-,

..22g

, 2--3, 242

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