Text

Amend 89 to License NPF-49,removing Listing, Enclosure Building Bypass Leakage Paths, from TSs & Making Numerous Editorial Changes
	ML20064N585
Person / Time
Site:	Millstone
Issue date:	03/24/1994
From:	Stolz J; Office of Nuclear Reactor Regulation
To:	; Northeast Nuclear Energy Co (NNECO)
Shared Package
ML20064N587	List: ML20064N585; ML20064N590;
References
	NPF-49-A-089 NUDOCS 9403300035
	Download: ML20064N585 (23)
	v • d • e

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.t UNITED STATES j

j NUCLEAR REGULATORY COMMISSION

't WASHINoTON, D.C. 206tE0001

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NORTHEAST NUCLEAR ENERGY COMPANY. ET AL.

DOCKET N0. 50-4L1 RILLSTONE NVCLEAR POWER STATION. UNIT NO. 3 AMEN 0 MENT TO FACILITY OPERATING LICEN_S_E Amendment No. 89 License No. NPF-49 1.

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

A.

The application for amendment by Northeast Nuclear Energy Company, et al. (the licensee), dated September 24, 1993, complies 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 Commission; C.

There is reasonable assurance (i) that the activities authorized by l

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; D.

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

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

PDR

. - ~

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-49 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 89

, and the Environmental Protection Plan contained in Appendix 8, both of which are attached hereto are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications and the l

Environmental Protection Plan.

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION C

S& t.c fr

//10 ohn F/ Stolz, Director Project Directorate I-4 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: March 24, 1994

-v

ATTACHMENT TO LICENSE AMENDMENT NO. 89 FACILITY OPERATING LICENSE N0. NPF-49 DOCKET N0. 50-423 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages.

The revised pages are identified by amendment number and contain vertical lines indicating the areas of change, Remove Insert I

ii 11 v

v viii viii ix ix xii xii xiii xiii xiv xiv j

xv xv l

l xviii xviii l

3/4 3-7 3/4 3-7 3/4~3-24 3/4 3-24 3/4 3-29 3/4 3-29 3/4 4-16 3/4 4-16 3/4 6-2 3/4 6-2 3/4 6-4 3/4 6-4 3/4 6-36a B3/4 4-11 83/4 4-11 B3/4 6-1 B3/4 6-1 B3/4 6-la B3/4 8-3 B3/4 8-3 B3/4 11-3 B3/4 11-3 1

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l INDEX DEFINITIONS SECTION IME 1.32 SLAVE RELAY TEST.............................................

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

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

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

1-6 1.36 TRIP ACTUATING DEVICE OPERATIONAL TEST.......................

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

1-6 1.38 UNRESTRICTED AREA............................................

1-6 J

1.39 VENTING......................................................

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

1-7 1.41 SPENT FUEL POOL STORAGE PATTERNS.............................

1-7 1.42 COREOPERATINGLIMITSREPORT(C0LR)..........................

1-7 l

ND 1.43 ALLOWED POWER LEVEL--APL 1,7 0l 1,44 ALLOWED POWER LEVEL--APL 1-7 1.45 THE CHARGING PUMP OPERABILITY................................

1-7 TABLE 1.1 FREQUENCY N0TATION......................................

1-8 i

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

1-9 MILLSTONE - UNIT 3 ii Amendment No. 19, J$ JS, M, 89 0139

~

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0VIREMENTS SECTION PME Position Indiction System -

Shutdown.....................

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

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

3/4 1-26 Control Rod Insertion Limits.............................

3/4 1-27 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE....................................

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

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

3/4 2-3 l

3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - F (Z).....................

3/4 2-5 Four Loops 0perating............g 3/4 2-5 Three Loops 0perating....................................

3/4 2-12 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACT0R...................................................

3/4 2-19 Four Loops 0perating.....................................

3/4 2-19 Three Loops 0perating....................................

3/4 2-22 3/4.2.4 QUADRANT POWER TILT RATI0................................

3/4 2-24 l

3/4.2.5 DNB PARAMETERS...........................................

3/4 2-27 l

TABLE 3.2-1 DNB PARAMETERS........................................

3/4 2-28 l

3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION......................

3/4 3-1 TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION...................

3/4 3-2 TABLE 3.3-2 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES....

3/4 3-8 TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS.............................................

3/4 3-10 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION..........................................

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

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

3/4 3 26 l

MILLSTONE - UNIT 3 v

Amendment No. JE, $9,89.

0042

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAqE l

1 FIGURE 3.4-1 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY >1pCi/ gram DOSE EQUIVALENT I-131..................

3/4 4-30 TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM.........................

3/4 4-31 3/4.4.9 PRESSURE / TEMPERATURE LIMITS...............

3/4 4-33 FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS -

APPLICABLE UP TO 10 EFPY 3/4 4-34 FIGURE 3.4-3 REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS -

APPLICABLE UP T0 10 EFPY 3/4 4-35 TABLE 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM -

WITHDRAWAL SCHEDULE...................

3/4 4-36 Pre s s uri zer.......................

3/4 4-3 7 j

Overpressure Protection Systems.............

3/4 4-38 FIGURE 3.4-4a NOMINAL MAXIMUM ALLOWABLE PORV SETPOINT FOR THE COLD 1

OVERPRESSURE SYSTEM (FOUR LOOP OPERATION) 3/4 4-40 FIGURE 3.4-4b NOMINAL MAXIMUM ALLOWABLE PORV SETPOINT FOR THE COLD OVERPRESSURE SYSTEM (THREE LOOP OPERATION).......

3/4 4-41 3/4.4.10 STRUCTURAL INTEGRITY 3/4 4-42 3/4.4.11 REACTOR COOLANT SYSTEM VENTS 3/4 4-43 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS 3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - T GREATER THAN OR EQUAL TO 350*F 3/4 5-3 j

3/4.5.3 ECCS SUBSYSTEMS - T, LESS THAN 350*F 3/4 5-7 3/4.5.4 REFUELING WATER STORAGE TANK..............

3/4 5-9 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Containment Integrity 3/4 6-1 Containment Leakage..................

3/4 6-2 Containment Air Locks 3/4 6-5 Containment Pressure..................

3/4 6-7 HILLSTONE - UNIT 3 viii Amendment No. A9, E/,89, 0199

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PEE Air Temperature 3/4 6-9 Containment Structural Integrity...........

3/4 6-10 Containment Ventilation System............

3/4 6-11 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Containment Quench Spray System 3/4 6-12 Recirculation Spray System..............

3/4 6-13 Spray Additive System 3/4 6-14 3/4.6.3 CONTAINMENT ISOLATION VALVES.............

3/4 6-15 3/4.6.4 COMBUSTIBLE GAS CONTROL Hydrogen Monitors 3/4 6-35 Electric Hydrogen Recombiners 3/4 6-36 3/4.6.5 SUBATMOSPHERIC PRESSURE CONTROL SYSTEM Steam Jet Air Ejector 3/4 6-37 3/4.6.6 SECONDARY CONTAINMENT Supplementary Leak Collection and Release System...

3/4 6-38 Secondary Containment Boundary............

3/4 6-41 Secondary Containment Boundary Structural Integrity.................

3/4 6-42 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE Safety Valves 3/4 7-1 TABLE 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING FOUR LOOP OPERATION 3/4 7-2 TABLE 3.7-2 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING THREE LOOP OPERATION 3/4 7-2 MILLSTONE - UNIT 3 ix Amendment No. A9,63, 87, 89, 0199

1 i

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.lRQG j

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS l

i SECTION PAG 1 3/4.9.6 REFUELING MACHINE........................................

3/4 9-6 3

i 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE AREAS..................

3/4 9-7 j

3/4.9.8.

RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION High Water Level.........................................

3/4 9-8

]

Low Water Level..........................................

3/4 9-9 3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM...........

3/4 9-10 3/4.9.10 WATER LEVEL - RE ACTOR VESSEL.............................

3/4 9-11 3/4.9.11 WATER LEVEL - STORAGE P0OL..............................

3/4 9-12 1

3/4.9.12 FUEL BUILDING EXHAUST FILTER SYSTEM......................

3/4 9-13 J

3/4.9.13 SPENT FUEL P00L -

REACTIVITY.............................

3/4 9-16 i

3/4.9.14 SPENT FUEL P0OL - STORAGE PATTERN........................

3/4 9-17 i

FIGURE 3.9-1 FUEL ASSEMBLY MINIMUM BURNUP VERSUS INITIAL U235 ENRICHMENT FOR STORAGE IN REGION II SPENT FUEL RACKS.,

3/4 9-18 j

FIGURE 3.9-2 REGION I THREE OF FOUR FUEL ASSEMBLY LOADING SCHEMATIC FOR A TYPICAL 6X6 STORAGE MODULE............

3/4 9-19 3/4.10 SPECIAL TEST EXCEPTIONS 1

3/4.10.1 SHUTDOWN MARGIN..........................................

3/4 10-1 j

3/4.10.2 GROUP HEIGHI, INSERTION, AND POWER DISTRIBUTION LIMITS

]

Four Loops Operating.....................................

3/4 10 2 j

Three Loops 0perating....................................

3/4 10-3 3/4.10.3 PHYSICS TESTS............................................

3/4 10-4 3/4.10.4 REACTOR COOLANT L00PS....................................

3/4 10-5 3/4.10.5 POSITION INDICATION SYSTEM -

SHUTD0WN....................

3/4 10-6 3/4.11 RADI0 ACTIVE EFFLUENTS

)

3/4.11.1 LIQUID EFFLUENTS Concentration............................................

3/4 11-1 Dose -

Liquids...........................................

3/4 11-2 i

3/4.11.2 GASE0US EFFLUENTS i

Dose Rate................................................

3/4 11-3 Dose - Noble' Gases.......................................

3/4 11-4 Dose - Radioiodines, Radioactive Material in Particulate Form and Radionuclides Other Than Noble Gases............

3/4 11-5 3/4.11.3 TOTAL 00SE...............................................

3/4 11-6 3

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MILLSTONE - UNIT 3 xii Amendment No. JS 89, 0141 i

F w

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1 7

meg w

w r'-1 w

ew u----w c-

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i J

INDEX 4

1 BASES J

j SECTION PEtE b

l 3/4.0 APPLICABILITY...............................................

B 3/4 0-1 j

l 3/4.1 REACTIVITY CONTROL SYSTEMS l

3/4.1.1 B0 RATION CONTR0L..........................................

B 3/4 1-1 j

3/4.1.2. BORATION SYSTEMS..........................................

B 3/4 1-2 i

3/4.1.3 MOVABLE CONTROL ASSEM8 LIES................................

B 3/4 1-3a l

4 1

(

1 3/4.2 POWER DISTRIBUTION LIMITS...................................

B 3/4 2-1 j

3/4.2.1 AXIAL FLUX DIFFERENCE.....................................

B 3/4 2-1 j

3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR.........

B 3/4 2-3 i

]

3/4.2.4 QUADRANT POWER TILT RATI0.................................

B 3/4 2-5 i

3/4.2.5 DNB PARAMETERS............................................

B 3/4 2-5 1

l t

1 3/4.3 INSTRUMENTATION i

3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM INSTRUMENTATION and i

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM l

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

B 3/4 3-1 l

3/4.3.3 MONITORING INSTRUMENTATION................................

B 3/4 3-3

]

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

B 3/4 3-6 l

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION.............

B3/44-1 3/4.4.2 SAFETY VALVES.............................................

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

B 3/4 4-2 j

3/4.4.4 RELIEF VALVES.............................................

B 3/4 4-2

^

3/4.4.5 STEAM GENERATORS..........................................

B 3/4 4-3 j

3/4.4.6 REACTOR COOLANT SYSTEM LEAXAGE............................

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

B 3/4 4-5, 3/4.4.8 SPECIFIC ACTIVITY.........................................

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

B 3/4 4-7 MILLSTONE - UNIT 3 xiii Amendment No. %g, 89 0141 1

,a

.,m_.

m

-.m._

.i._.

m m

INDEX BASES SECTION EME I

4 TABLE B 3/4.4-1 REACTOR VESSEL FRACTURE TOUGHNESS PROPERTIES......

B 3/4 4-9 FIGURE B 3/4.4-1 FAST NEUTRON FLUENCE (E>1MeV) AS A FUNCTION OF l

FULL POWER SERVICE LIFE..................................

B 3/4 4-10 3/4.4.10 STRUCTURAL INTEGRITY.....................................

B 3/4 4-15 3/4.4.11 REACTOR COOLANT SYSTEM VENTS.............................

B 3/4 4-15 l

3/4.5 EMERGENCY CORE COOLING SYSTEMS 1

3/4.5.1 ACCUMULATORS..............................................

B 3/4 5-1 i

3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS...............................

B 3/4 5-1 3/4.5.4 REFUELING WATER STORAGE TANK..............................

B 3/4 5-2 a

1 1

l 3/4.6 CONTAINMENT SYSTEMS i

l 3/4.6.1 PRIMARY CONTAINMENT.......................................

B 3/4 6-1 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS......................

B 3/4 6-2

]

3/4.6.3 CONTAINMENT ISOLATION VALVES..............................

B 3/4 6-3

]

3/4.6.4 COMBUSTIBLE GAS CONTR0L...................................

B 3/4 6-3 3/4.6.5 SUBATMOSPHERIC PRESSURE CONTROL SYSTEM....................

B 3/4 6-3b l 3/4.6.6 SECONDARY CONTAINMENT.....................................

B 3/4 6-4 d

3/4.7 PLANT SYSTEMS i

l 3/4.7.1 TURBINE CYCLE.............................................

B 3/4 7-1 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION...........

B 3/4 7-3 3/4.7.3 REACTOR PLANT COMPONENT COOLING WATER SYSTEM..............

B 3/4 7-3 3/4.7.4 SERVICE WATER SYSTEM......................................

B 3/4 7-3 t

3/4.7.5 ULTIMATE HEAT SINK........................................

B 3/4 7-3 I

3/4.7.6 FLOOD PROTECTION..........................................

B 3/4 7-4 1

3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM.................

B 3/4 7-4 3/4.7.8 CONTROL ROOM ENVELOPE PRESSURIZATION SYSTEM...............

B 3/4 7-4 3/4.7.9 AUXILIARY BUILDING FILTER SYSTEM..........................

B 3/4 7-4 2

3/4.7.10 SNUBBERS..................................................

B 3/4 7-5 MILLSTONE - UNIT 3 xiv Amendment No. g, 89, 0141

INDEX BASES SECTION PME 3/4.7.11 SEALED SOURCE CONTAMINATION B 3/4 7-6 3/4.7.12 DELETED 3/4.7.13 DELETED 3/4.7.14 AREA TEMPERATURE MONITORING,..............

B 3/4 7-8 3/4.8 ELECTRICAL POWER SYSTEtil 3/4.8.1, 3/4.8.2, and 3/4.8.3 A.C. SOURCES, D.C. SOURCES, AND ONSITE POWER DISTRIBUTION B 3/4 8-1 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES B 3/4 8-3 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION...................

B 3/4 9-1 4

3/4.9.2 INSTRUMENTATION.....................

B 3/4 9-1 i

3/4.9.3 DECAY TIME B 3/4 9-1 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS............

B 3/4 9-1 3/4.9.5 COMMUNICATIONS B 3/4 9-1 3/4.9.6 REFUELING MACHINE....................

B 3/4 9-2 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE AREAS B 3/4 9-2 3/4.9.8 RESIDUAL HEAT REMOVAL AND C0OLANT CIRCULATION B 3/4 9-2 3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM B 3/4 9-2 3/4.9.10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL AND STORAGE P0OL B 3/4 9-3 3/4.9.12 FUEL BUILDING EXHAUST FILTER SYSTEM...........

B 3/4 9-3 3/4.9.13 SPENT FUEL POOL - REACTIVITY B 3/4 9-3 3/4.9.14 SPENT FUEL P0OL - STORAGE PATTERN B 34/ 9-3 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN..................... B 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS

. B 3/4 10-l 3/4.10.3 PHYSICS TESTS...................... B 3/4 10-1 3/4.10.4 REACTOR COOLANT LOOPS.................. B 3/4 10-1 3/4.10.5 POSITION INDICATION SYSTEM - SHUTDOWN.......... B 3/4 10-1 MILLSTONE - UNIT 3 xv Amendment No. EA, 89, 0200

l 1

INDEX ADMINISTRATIVE CONTROLS SECTION E8E I

1 6.1 RESPONSIBILITY...................................................

6-1 6.2 ORGANIZATION....................................................

6-1 l

6.2.1 0FFSITE AND'ONSITE ORGANIZATIONS...............................

6-1 l

t f,

6.2.2 FACILITY STAFF................................................ 6-1

}

i j

TABLE 6.2-1 MINIMUM SHIFT CREW COMPOSITION.......................... 6-3 6.2.3 INDEPENDENT SAFETY ENGINEERING GROUP (ISEG)

Function......................................................

6-4 Composition...................................................

6-4 Responsibilities.............................................. 6-4 Records.......................................................

6-4 6.2.4 SHIFT TECHNICAL ADVISOR.......................................

6-4 6.3 FACILITY STAFF OUALIFICATIONS................................... 6-5 6.4 TRAINING........................................................

6-5 6.5 REVIEW AND AUDIT................................................

6-5 i

'I 6.5.1 PLANT OPERATIONS REVIEW COMMITTEE (PORC)

Function...................................................... 6-5 Composition...................................................

6-5 Al t e r n a t e s....................................................

6 - 6 Meeting Frequency.............................................

6-6 Quorum........................................................ 6-6 Re s p o n s i b i l i t i e s..............................................

6 - 6 Authority.....................................................

6-7 Records.......................................................

6-7 MILLSTONE - UNIT 3 xviii Amendment No. JJ, %),89, 0142

=

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

TABLE 3.3-1 (Continued)

}

ACTION STATEMENTS (Continued) j*

ACTION 9 -

(Notused) l

)

ACTION 10 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY j

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 i

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 i

Channels OPERABLE requirement, restore the inoperable channel j

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.

1 ACTION 12 - With the number of 0PERABLE channels one less than the Total i

Number of Channels, STARTUP and/or POWER OPERATION may proceed i

provided the following conditions are satisfied:

a.

The inoperable channel is placed in the tripped condition 1

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

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 i

I surveillance testing of the Turbine. control valves.

j ACTION 13 - With one of the diverse trip features (undervoltage or shunt-trip attachments) inoperable, restore it to OPERABLE status i

within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or declare the breaker inoperable and apply i

ACTION 10.

The breaker shall not be bypassed while one of the j

diverse ' trip features. is inoperable except for the time required for performing maintenance to restore the breaker to 1

OPERABLE status, t

i ACTION 13A - With the number of OPERABLE channels one less than the Minimum j

Channels OPERABLE requirement, restore the inoperable Channel 4

to OPERABLE status within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 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 ; however, one channel may be bypassed i

for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing per Specification 4.3.1.1, provided the other channel is operable.

j.

1 11 4

MILLSTONE - UNIT 3 3/4 3-7 Amendment No. 7), 89, j

ooss

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-11 (Pressurizer Pressure Interlock) Setpoint.

- - - Trip function automatically blocked above P-11 and may be blocked below P-11 when Safety Injection on low steam line pressure is not blocked.

ACTION STATEMENTS i

ACTION 14 - 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 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 ; however, one 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 per Specification 4.3.2.1, provided the other channel is OPERABLE.

ACTION 15 - (not used).

ACTION 16 - 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 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

,4 ACTION 17 - 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 r

be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing per Specification 4.3.2.1.

ACTION 18 - With less than the Minimum Channels OPERABLE requirement, within I hour initiate and maintain operation of the Control Room Emergency Ventilation System in the recirculation mode of 1

operation.

MILLSTONE - UNIT 3 3/4 3-24 Amendment No. J57, 7A 89, 0063 l

TABLE 3.3-4 (Continued)

Si'i ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS

"F$

SENSOR e

TOTAL ERROR FUNCTIONAL UNIT ALLOWANCE fTA) I fS)

TRIP SETPOINT ALLOWABLE VALUE E

6.

Auxiliary Feedwater (Continued)

Z

2) Start Turbine-18.10 16.64 1.50 2 18.10% of 1 17.11% of narrow w

Driven Pumps narrow range range instrument instrument span.

span.

d.

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

e.

Loss-of-Offsite Power N.A.

N.A.

1 2800V 1 2720V Start Motor-Driven Pumps w1 f.

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

w A

Actuation (CDA) Start Motor-Driven Pumps 7.

Control Building Isolation 4

a.

Manual Actuation N.A.

N.A.

N.A N.A.

b.

Manual Safety Injection N.A.

N.A.

N.A N.A.

{

g Actuation E.

R c.

Automatic Actuation N.A.

N.A.

Logic and Actuation 4

Relays 2

i

?

d.

Containment 3.3 1.01 1.75 5 3.0 psig 5 3.8 psig w?

Pressure--High 1 e.

Control Building N.A.

N.A.

N.A 51.5 x 10 pci/cc

$1.5 x 10 pci/ccl

-5

-5 g

Inlet Ventilation Radiation

REACTOR COOLANT SYSTEM STEAM GENERATORS SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.3 Insoection Frecuencies - The above required inservice inspections of steam generator tubes shall be performed at the following frequencies:

)

a.

Inservice inspections shall be performed at intervals of not less than l

12 nor more than 24 calendar months

after the previous inspection.

If two consecutive inspections, not including the areservice inspection, result in all inspection results falling into tie C-1 category or if two consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months; b.

If the results of the inservice inspection of a steam generator conducted in accordance with Table 4.4-2 at 40-month intervals fall in Category C-3, the inspection frequency shall be increased to at least once per 20 months.

The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria of Specification 4.4.5.3a.; the interval may then be extended to a maximum of once per 40 months; and c.

Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection specified in Table 4.4-2 during the shutdown subsequent to any of the following conditions:

1)

Primary-to-secondary tubes leak (not including leaks i

originating from tube-to tube sheet welds) in excess of the limits of Specification 3.4.6.2, or 2)

A seismic occurrence greater than the Operating Basis Earthquake, or 3)

A loss-of-coolant accident requiring actuation of the Engineered Safety Features, or 4)

A main steam line or feedwater line break.

Except that the surveillance related to the steam generator inspection, due no later than August 21, 1993, may be deferred until the next refueling outage or no later than September 30, 1993, whichever is earlier.

MILLSTONE - UNIT 3 3/4 4-16 Amendment No. 57,89 0146

CANTAINNENT SYSTENS CONTAINNENT LEMAAE LINITING CONDITION FOR OPERATION 3.6.1.2 Containment leakage rates shall be limited to:

a.

An overall integrated leakage rate of less than or equal to L.,

0.3% by weight of the f.ontainment air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at P.,

53.27 psia (38.57 psig);

b.

A combined leakage rate of less than 0.60 L, for all penetrations and valves subject to Type B and C tests, when pressurized to P,;

and c.

A combined leakage rate of less than or equal to 0.042 L, for all penetrations that are SECONDARY CONTAINMENT B0VHDARY bypass leakage paths when pressurized to P,.

APPLICABILITY: H00ES 1, 2, 3, and 4.

ACHOB:

With the measured overall integrated containment leakage rate exceeding 0.75 L, or the measured combined leakage rate for all penetrations and valves subject to Type B and C tests exceeding 0.60 L., or the combined bypass leakage rate exceeding 0.042 L., restore the overall integrated leakage rate to less than 0.75 L, the combined leakage rate for all penetrations subject to Type B and C tests to less than 0.60 L,, and the combined bypass leakage rate to less than 0.042 L, prior to increasing the Reactor Coolant System temperature above 200'F.

SURVEILLANCE REQUIRENENTS 4.6.1.2 The containment leakage rates shall be demonstrated at the following test schedule and shall be determined in conformance with the criteria specified in Appendix J of 10 CFR Part 50 using methods and provisions of ANSI N45.4-1972 (Total Time Method) and/or ANSI /ANS 56.8 1981 (Mass Point Method):

a.

Three Type A tests (Overall Integrated Containment Leakage Rate)

h:ll be conducted at 40 i 10 month intervals during shutdown at a pressure not less than P., 53.27 psia (38.57 psig) during each 10-year service period. The third test of each set shall be conducted during the shutdown for the 10 year plant inservice inspection; b.

If any periodic Type A test fails to meet 0.75 L, the test schedule for subsequent Type A tests shall be reviewed and approved by the Comission, if two consecutive Type A tests fail to meet 0.75 L,,

a Type A test shall be performed at least every 18 months until two consecutive Type A tests meet 0.75 L, at which time the above test schedule may be resumed; HILLSTONE - UNIT 3 3/4 6-2 Amendment No. A9, E/, 89, 0201

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i MILLSTONE - UNIT 3 3/4 6-4 Amendment No. JJ, EJ 89, j

O202 i

I

k*

REACTOR COOLANT SYSTEM l

,)-

BASES t

l PRESSURE / TEMPERATURE LIMITS (Continued)

Values of ART determined in this manner may be used until the results from the material Nve111ance program, evaluated according to ASTM E185, are i

j available.

Capsules will be removed in accordance with the requirements of j

ASTM E185-73 and 10 CFR Part 50, Appendix H.

The surveillance specimen with-drawal schedule is shown in Table 4.4-5.

The lead factor represents the rela-j tionship between the fast neutron flux density at the' location of the capsule and the inner wall of the reactor vessel.

Therefore, the results obtained i

from the surveillance specimens can be used to predict future radiation damage i

to the reactor vessel material by using the lead factor and the withdrawal

{

time of the capsule. The heatup and cooldown curves must be recalculated when the ART determined from the surveillance capsule exceeds the calculated NDT l

ART for the equivalent capsule radiation exposure, j

NDT Allowable pressure-temperature relationships for various heatup and R

cooldown rates are calculated using methods derived from Appendix G in Sec-i tion III of the ASME Boiler and Pressure Vessel Code _as required by j

Appendix G to 10 CFR Part 50, and these methods are discussed in detail in the following paragraphs.

i The general method for calculating heatup and cooldown limit curves is based upon the principles of the linear elastic fracture mechanics (LEFM) technology.

In the calculation procedures a semielliptical surface defect with a depth of one-quarter of the wall thickness, T. and a length of 3/2T is assumed to exist at the inside of the vessel wa;l as well as at the outside of the vessel wall.

The dimensions of this postulated crack, referred to in Appendix G of ASME Section I?I as the reference flaw, amply t

exceed the current capabilities of in:,ervice ' inspection techniques.

Therefore, the reactor operation limit curves developed for this reference crack are conservative and provide sufficient safety margins for protection against nonductile failure.

To assure that the radiation embrittlement l

effects are accounted for in the calculation of the limit curves, the most limiting value of the nil-ductility reference temperature, RT is used

NDT, and this includes the radiation-induced shift, ART corresponding to the end of the period for which heatup and cooldown curv$T,re generated.

a The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, K, for the combined thermal and pressure stresses at any time during heatup j

or cooldown cannot be greater than the reference stress intensity factor, K for the metal temperature at that time.

K is obtained from the refered$e, IR fracture toughness curve, defined in Appendix.G to the ASME Code.

The K IR curve is given by the equation:

MILLSTONE - UNIT 3 B 3/4 4-11 Amendment No. f#, 89, 0147 i

o

,-y y.

y.

y..,

...m_..,,,.,,,..m,...,_g u m.

3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 CONTAINMENT INTEGRITY Primary CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the safety analyses. This restric-tion, in conjunction with the leakage rate limitation, will limit the SITE B0UNDARY radiation doses to within the dose guidelines of 10 CFR Part 100 during accident conditions and the control room operators dose to within the guidelines of GOC 19.

l 3/4.6.1.2 CONTAINMENT LEAKAGE The limitations on containment leakage rates ensure that the total containment leakage volume will not exceed the value assumed in the safety analyses at the peak accident pressure, P,.

As an added conservatism, the measured overall integrated leakage rate is further limited to less than or equal to 0.75 L.

during performance of the periodic test to account for possible degradation of the containment leakage barriers between leakage tests.

The surveillance testing for measuring leakage rates are consistent with the requirements of Appendix J of 10 CFR Part 50.

The enclosure building bypass leakage paths are listed in Operating Procedure 3273, " Technical Requirements - Supplementary Technical Specifica-tions."

The addition or deletion of the enclosure buildini bypass leakage paths shall be made in accordance with Section 50.59 of 10CFF50 and approved by the Plant Operation Review Committee.

3/4.6.1.3 CONTAINMENT AIR LOCKS The limitations on closure and leak rate for the containment air locks are required to meet the restrictions on CONTAINMENT INTEGRITY and containment leak rate.

Surveillance testing of the air lock seals provides assurance that the overall air lock leakage will not become excessive due to seal damage during the intervals between air lock leakage tests.

3/4.6.1.4 and 3/4.6.1.5 AIR PRESSVRE and AIR TEMPERATVE The limitations on containment pressure and average air temperature ensure that:

(1) the containment structure is prevented from exceeding its design negative pressure of 8 psia, and (2) the containment peak pressure does MILLSTONE - UNIT 3 B 3/4 6-1 Amendment No. 57, 89, 0148

i I-4 j-3/4.6 CONTAINMENT SYSTEMS i

BASES 3/4.6.1.4 and 3/4.6.1.5 AIR PRESSVRE and AIR TEMPERATURE (continued) l l

not exceed the design pressure of 60 psia during LOCA conditions.

Measure-ments shall be made at all listed locations, whether by fixed or portable instruments, prior to determining the average air' temperature.

The limits on the pressure and average air temperature are consistent with the assumptions 1

l of the safety analysis.

The minimum total containment pressure of 10.6 psia is determined by summing the minimum permissible air partial pressure of j

8.9 psia and the maximum expected vapor pressure of 1.7 psia (occurring at the maximum permissible containment initial temperature of 120*F).

i 1

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1 MILLSTONE - UNIT 3 B 3/4 6-la Amendment No. JJ 89, 0148 3

i j

i

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l i

l ELECTRICAL POWER SYSTEMS BASES i

3/4.8.4 ELECTRICAL E0VIPMENT PROTECTIVE DEVICES Containment electrical penetrations and penetration conductors are pro-tected by either oeenergizing circuits not required during reactor operation or by demonstrating the OPERABILITY of primary and backup overcurrent protec-tion circuit breakers during periodic surveillance.

The Survaillance Requirements applicable to lower voltage circuit breakers provide as:urance of breaker reliability by testing at least one representative sample of each manufacturer's brand of circuit breaker.

Each manufacturer's l

molded case and metal case circuit breakers are grouped into representative l

samples which are then tested on a rotating basis to ensure that all breakers are tested.

If a wide variety exists within any manufacturer's brand of circuit breakers, it is necessary to divide that manufacturer's bieakers into groups and treat each group as a separate type of breaker for surveillance i

purposes.

l The molded case circuit breakers and unitized starters will be tested in i

accordance with Manufacturer's Instructions.

The OPERABILITY of the motor-operated valves thermal overload protection and integral bypass devices ensures that the thermal overload protection will not prevent safety-related valves from performing their function. The Surveil-lance Requirements for demonstrating the OPERABILITY of the thermal overload protection are in accordance with Regulatory Guide 1.106, " Thermal Overload Protection for Electric Hotors on Motor Operated Valves,"-Revision 1, March 1977.

Supplementary l

Operating Procedure

3273,

" Technical Requirements Technical Specifications," list containment penetration conductor overcurrent j

protective devices and thermal overload protection bypassed only under accident conditions and thermal overload protection not bypassed under accident conditions.

The addition or deletion of any device shall be made in accordance with Section 50.59 of 10CFR50 and approved by the Plant Operation Review Comittee.

MILLSTONE - UNIT 3 8 3/4 8-3 Amendment No. M 89, 0149

RADIOACTIVE EFFLUENTS BASES DOSE - RADI0 IODINES. RADI0 ACTIVE MATERIAL IN PARTICULATE FORM AND RADIONUCLIDES OTHER THAN NOBLE GASES (Continued) that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." The REMODCM calculational methods specified in the Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calcula-tional procedures based on models and data such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially under-estimated. The REM 0DCM calculational methodology and parameters for calculating the doses due to the actual release rates of the subject materials are consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evalu-ating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and j

Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Disper-sion of Gaseous Effluents in Routine Releases from Light-Water-cooled Reactors,"

Revision 1, July 1977. The release rate specifications for radiciodines and i

radionuclides in particulate form and radionuclides other than noble gases are i

dependent upon the existing radionuclide pathways to man.

The pathways that are examined in the development of these calculations are:

(1) individual inhalation of airborne radionuclides, (2) deposition of radionuclides onto green leafy vege-4 tation with subsequent consumption by man, (3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the milk and j

meat by man, and (4) deposition on the ground with subsequent exposure to man.

]

3/4.11.3 TOTAL DOSE l

This specification is provided to meet the dose limitations of 40 CFR Part 190.

For the purposes of the Special Report, it may be assumed that the dose commitment to any REAL MEMBER OF THE PUBLIC from other uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 5 miles must be considered.

MILLSTONE - UNIT 3 B 3/4 11-3 Amendment No.89, 0150

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

ML20064N585

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