ML20056C078

Proposed Tech Specs Changes,Incorporating Editorial Cleanup

ML20056C078
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	03/16/1993
From:	CONNECTICUT YANKEE ATOMIC POWER CO.
To:
Shared Package
ML20056C077	List: B14369, Application for Amend to License DPR-61,changing TS Re Editorial Cleanup,Including Incorporation of Missing Sections in Index,Providing Editorial Consistency Throughout Document & Removal of Cycle Specific Comments ML20056C078
References
NUDOCS 9303290014
Download: ML20056C078 (108)

Text

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Docket No. 50-213 B14369

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f Haddam Neck Plant Marked Up Proposed Revision to Technical Specifications Editorial Cleanup of Technical Specifications i

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March 1993 9303290014 930316 PDR ADOCK 05000213 P

PDR

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l July 19, 1990 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS j

l J

ff.GI SECTION s

3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems...............................

3/4 4-29

)

Ope rat i on al Le aka g e.....................................

3/4 4-31.

f 3/4.4.7 CHEMISTRY...............................................

3/4 4-33 TABLE 3.4-1 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS..............

3/4 4-34 r

TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS............................

3/4 4-35 i

3/4.4.8 SPECIFIC ACTIVITY.......................................

3/4 4-36 t

FIGURE 3.4-2 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC l

L ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY GREATER l

THAN I microcurte/ gram DOSE EQUIVALENT I-131........

3/4 4-37 TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM.................................

3/4 4-38 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Cool ant System..................................

3/4 4-39 j

FIGURE 3.4-3 CONNECTICUT YANKEE LIMIT CURVE FOR HYDROSTATIC AND l

i LEAK TESTING APPLICABLE FOR 22.0 EFFECTIVE FULL POWER YEARS.........................................

3/4 4-41 l

FIGURE 3.4-4 CONNECTICUT YANKEE REACTOR COOLANT SYSTEM HEATUP j

LIMITATIONS FOR 22.0 EFFECTIVE FULL POWER YEARS.....

3/4 4-42 i

i FIGURE 3.4-5 CONNECTICUT YANKEE REACTOR COOLANT SYSTEM C00LDOWN t

5 LIMITATIONS FOR 22.0 EFFECTIVE FULL' POWER YEARS.....

3/44-43 g_

l

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Pressurizer..........................................

3/4 4-45 j

Low Temperature Overpressure Protection Systems......

3/4 4-46 j

3/4.4.10 STRUCTURAL INTEGRITY....................................

3/4 4-48

'I 3/4.4.11 REACTOR COOLANT SYSTEM VENTS............................

3/4 4-49 3/4 + s1 314.Wil Faruch FwetMs l

i f

1 l

l HADDAM NECK VIII Amendment No. 175. 128 l

NYtt 4 D ON INDEX LIMITING CGFDlIIONS FOR OPERATION AND SURVElttANCE REOUIREMENTS EAGE l

SECTION 3 /4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ECCS SUBSYSTEMS - Tavg GREATER TRAN OR EQUAL TO 350 F...

3/4 5-1 TABLE 4.5-1 SAFETY INJECTION ACTUATED AUTOMATIC VALVES...........3/4 5-6 ECCS SUBSYSTEMS - Tavg LESS THAN 350'F..................

3/4 5-7 3/4.5.2 3/4.5.3 REFUELING WATER STORAGE T ANK............................

3/4 5-9 3/4 5-10 3/4.5.4 pH CONTROL SYSTEM.......................................

3 /4. 6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT 3/4 6-1 Containment Integrity...................................

3/4 6-2 Containment Leakage.....................................

3/4 6-4 Containment Air Locks...................................

3/4 6-6 l

Internt(Pressure.......................................

3/4 6-7 Air Temptcature.........................................

Containment Ve s sel St ructural Integrity.................

3/4 6-8 Containment Ventil ation System..........................

3/4 6-9 3/4.6.2 CONT AINMENT AIR RECIRCULATION SYSTEM....................

3/4 6-10 3/4.6.3 CONTAINMENT ISOLATION VALVES............................

3/4 6-12 3/4.7 PLANTS SYSTEMS 3/4.7.1 TURBINE CYCLE 3/4 7-1 l

Safety Va1ves...........................................

TABLE 3.7-1 STEAM LINE SAFETY VALVES PER L00P....................

3/47-2 3/4 7-3 Auxiliary Feedwater System..............................

3/4 7-4 Auxiliary Feedwater Supp1y..............................

3/4 7-5 Specific Activity.......................................

• M

-Tantz 4.5-2. Ec45 Manual Wcues i

HADDAM NECK IX Amendment No. 125

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Nay 8, 1992 INDEX LIMITING CONDITIONS FOR OPEE HIAN AND SURVEILLANCE RE00_1REMENTS ff_g SECTION TABLE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM..........................

3/4 7-6 Main Steam Line Trip Va1ves.............................

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

3/4 7-8 3/47-9 3/4.7.3 S ERV IC E W AT E R SYST EM....................................

3/4 7-10 3/4.7.4 SNUBBERS................................................

TABLE 4.7-2 SNUBBER VISUAL INSPECTION INTERVAL...................

3/4 7-10a 3/4.7.5 SEALED SOURCE CONTAMINATION.............................

3/4 7-14 3/4.7.6 FIRE SUPPRESSION SYSTEMS Fire Water Supply / Distribution System...................

3/4 7-16 Spray and/or Spri nkl er Systems..........................

3/4 7-19 3/4 7-21 C0 Systems.............................................

2 3/4 7-22 Halon Systems...

Fire. Hose Stations.....................................

3/4 7-23 3/4 7-24 TABLE 3.7-4 FIRE HOSE STATIONS.............

Yard Fire Hydrants and Associated u.,m..m Hose Houses...

3/4 7-25 TABLE 3.7-5 YARD FIRE HYDRANTS...................................

3/47-27 3/4.7.7 FIRE RATED ASSEMBLIES...................................

3/4 7-28 l

3/4.7.8 FLAMMABLE LIQUIDS CONTR0L...............................

3/4 7-30 3/4.7.9 FEEDWATER ISOLATION VALVES..............................

3/4 7-31 TABLE 3.7-6 FEEDWATER ISOLATION VALVES...........................

3/4 7-32 3/4.7.10 EXTERNAL FLOOD PROTECTION...............................

3/4 7-33 3/4.7.11 PRIMARY AUXILIARY BUILDING AIR CLEANUP SYSTEM........... 3/4 7-34 3/4.8 ELECTRICAL POWER SYSTEMS i

3/4.8.1 A.C. SOURCES j

3/4 8-1 Operating...............................................

TABLE 4.8-1 DI ESEL GENERATOR TEST SCHEDULE.......................

3/4 8-6 Shutdown................................................

3/48-7 3/4.8.2 D.C. SOURCES Operating...............................................

3/4 8-8 TABLE 4.8-2 BATTERY SURVEILLANCE REQUIREMENTS....................

3/4 8-10 Shutdown................................................

3/4 8-11 HADDAM NECK X

Amendment No. J7E,151 enn

o INDEX July 9, 1990 LIMITING CONDITIONS FOR OPERATION AQSUPVEillAELE[OUIREMENTS ff.GI SECTION 3/4.8.3 ONSITE POWER DISTRIBUTION 3/4 8-12 Operating...............................................

3/4 8-14 Shutdown................................................

3/4.9 REFUELING OPERATIONS 3/4 9-1 3/4.9.1 BORON CONC ENT RAT I ON.....................................

3/4 9-2 3/4.9.2 INSTRUMENTATION.........................................

3/4 9-3 3/4.9.3 DECAY TIME..............................................

3/4.9.4 CONTAINMENT BUILDING PENETRATIONS.......................

3/4 9-4 3/4 9-S 3/4.9.5 COMMUN I C AT I ON S..........................................

3/4 9-6 3/4.9.6 KANI PUL ATOR C RAN E...................

3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE 00 BUILDING.........

3/4 9-7 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION 3/4 9-8 I

High Water Leve1........................................

3/4 9-9 Low Water Leve1.........................................

5 3/4.9.9 CONTAINMENT PURGE SUPPLY, PURGE EXHAUST, AND PURGE EXHAUST BYPASS ISOLATION SYSTEM...................

3/4 9-10 WATER LEVEL - REACTOR VESSEL........................~....

3c 4, 9-11 3/4 9.10 3/4 9-12 3/4.9.11 WATER LEVEL-STORAGEP00L................................

3/4.9.12 FUEL STORAGE BUILDING AIR CLEANUP SYSTEM................

3/4 9-13 3/4.10 SPECIAL TEST EXCEPTIONS 3/4 10-1 3/4.10.1 SHUTDOWN KARGIN..........

3/4 10-2 3/4.10.2 PHYS I CS T E S T S...........................................

3/4.10.3 POSITION INDICATION SYSTEM - SHUTD0WN...................

3/4'10-3 3/4.10.4 POSITION INDICATION SYSTEM - OPERATING..................

3/4 10-4 l

q 3/4.11 RID 10 ACTIVE EFFLUENTS 1

3/4.11.1 LIQUID EFFLUENTS 3/4 11-1 Concentration...........................................

s 3/4 11-2

Dose, Liquids...........................................

HADDAM NECK XI Amendment No. )?( 127 1

J

r APR 2 61990 4

INDEX BASES PAGE SECTION B 3/4 0-1 3/4.0 APPLICABILITY...........................................

3/4.1 REACTIVITY CONTROL SYSTEMS B 3/4 1-1 3/4.1.1 BORATION CONTR0L........................................

B 3/4 1-2 3/4.1.2 B0 RATION SYSTEMS........................................

B 3/4 1-3 3/4.1.3 MOVABLE CONTROL ASSEMBLIES..............................

}/.4 2 POWER DISTRIBUTION LIMITS s

B 3/4 2-1 3/4.2.1 AXIAL 0FFSET............................................

B 3/4 2-1 3/4.2.2 LINEAR HEAT GENERATION RATE.............................

NUCLEARENTHALPYRISEHOTCHANNELFACTOR.fb...........

B 3/4 2-1 3/4.2.3 B3/42ff 3/4.2.4 QU ADRANT POWER T ILT RAT I0...............................

B 3/4 2(7 1 3/4.2.5 DNB PARAMETERS..........................................

3/4.3 INSTRUMENTATION 3/4.3.1 & 3/4.3.2 REACTOR TRIP SYSTEM INSTRUMENTATION AND ENGINE SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION........B 3/4 3-1 B 3/4 3-2 3/4.3.3 MONITORING INSTRUMENTATION..............................

B3/43@

3/4.3.4,FLOODPSi. PROTECTION.....................................

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

B 3/4 4-1 B 3/4 4-2 3/4.4.2 SAFETY VALVES..........................................

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

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

B 3/4 4-3 3/4.4.5 ST E AM G EN E RAT ORS.......................................

3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE.........................

B 3/4 4-5 HADDAM NECK XIII Amendment No. 125

e APR 261950 I

BASES EA.E SECTION B 3/4 4-7 l

3/4.4.7 CHEMISTRY..............................................

B 3/4 4-7 j

3/4.4.8 SPECIFIC ACTIVITY......................................

B 3/4 4-8 3/4.4.9 PRESSURE / TEMPERATURE L1MITS............................

B3/44-12 3/4.4.10 STRUCTURAL INTEGRITY...................................

B 3/4 4-12 l

3/4.4.11 REACTOR COOLANT SYSTEM VENTS...........................

B V4 4-2 l

3N.Y.it Fmt.E0 Fust. bs 3/4.5 EMERGENCY CORE COOLING SYSTEMS B 3/4 5-1 l

3/4.5.1 & 3/4.5.2 ECCS SUBSYSTEMS................................

B 3/4 5-2 3/4.5.3 REFUELING WATER STORAGE TANK...........................

B 3/4 5-2' 3/4.5.4 pH CONTROL SYSTEM......................................

c s

3/4.6 CONTAINMENT SYSTEMS B 3/4 6-1 3/4.6.1 PRIMARY CONTAINMENT.....................................

B 3/4 6-3 3/4.6.2 CONTAINMENT AIR RECIRCULATION SYSTEM....................

1 B 3/4 6-3 3/4.6.3 CONTAINMEFT ISOLATION VALVES............................

)

3/4.7 PLANT SYSTEMS l

B 3/4 7-1 l

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

3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION.........

B 3/4_7-2 l

B 3/4 7-2

-l 3/4.7.3 SERVICE WATER SYSTEM...................................

B 3/4 7-3 3/4.7.4 SNUBBERS................................................

B 3/4 7-4 l

'3/4.7.5 SEALED SOURCE CONTAMINATION.............................

1 B 3/4 7-4 j

3/4.7.6 FIRE-SUPPRESSION SYSTEMS................................

B 3/4 7-5 j

3/4.7.7 F I RE RAT ED ASSEMBL I E S...................................

i B 3/4 7-5 3/4.7.B FLAMMABLE LIQUIDS CONTR0L...............................

I B 3/4 7-6 3/4.7.9 F EEDWATER ISOLATION VALVES..............................

B 3/4 7-6 3/4.7.10 EXTERNAL F LOOD PROTECTION...............................

B 3/4 7-6 3/4.7.11 PRIMARY AUXILIARY BullDING AIR CLEANUP SYSTEM...........

HADDAM NECK XIV Amendment No. 125

r 1

July 9, 1990 l

,[glX I

j PSES I

U i

SECTION b

3/4.8 ELECTRICAL POWER SYSTEMS y

su6 sl4.8.3 l

A.C. SOURCES, D.C. SOURCES, ONSITE POWER 3/4.8.1;r 3/4.8.2 B 3/4 8-1 D I ST RI BUT I ON............................................

3/4.9 REFUELING OPERATIONS B 3/4 9-1 3/4.9.1 BDRON CONCENTRAT I ON.....................................

B 3/4 9-1 j

3/4.9.2 I NST RUMENT AT I ON.........................................

B 3/4 9-1 3/4.9.3 DECAY TIME...............................................

B 3/4 9-1 3/4.9.4 CONTAINMENT BUI LDING PENET RATIONS.......................

B 3/4 9-1 3/4.9.5 COMMUN I CAT I ON S..........................................

B 3/4 9-2 3/4.9.6 MANIPULATOR CRANE.......................................

B 3/4 9-2 f

CRANE TRAVEL - SPENT FUEL STORAGE' BUILDING..............

3/4.9.7 B 3/4 9-2 RESIDUAL HEAT REM 3 VAL AND COOLANT CIRCULATION...........

l 3/4.9.8 CONTAINMENT PURGE SUPPLY, PURGE EXHAUST, AND B 3/4 9-2 f

3/4.9.9 PURGE EXHAUST BYPASS ISOLATION SYSTEM...................

WATER LEVEL - REACTOR VESSEL AND STORAGE 3/4.9.10 & 3/4.9.11 B 3/4 9-3 P00L....................................................

B 3/4 9-3 FUEL STORAGE BUILDING AIR CLEANUP SYSTEM................

3/4.9.12 3 /A.10 SPEC 1A1 TEST EXCEPTIONS B 3/4 10-1

,. 10.1 SHUTDOWN KARGIN.........................................

B 3/4 10-1 j

.10.2 PHY S I C S T E ST S........................................... '

B3/4.30-1 POSITION INDICATION SYSTEM-SHUTDOWN.....................

B 3/4 10-1 l

3/4.10.3 POSITION INDICATION SYSTEM - OPERATING..................

'3/4.10.4 3/4.11 RADIDACTIVE EFFLUENTS B 3/4 ll'-1 3/4.11.1 LIQUID EFFLUENTS........................................

B 3/4 11 ;

3/4.11.2 GASEOUS EFFLUENTS.......................................

B 3/4 11-3 l

3/4.11.3 TOTAL D0SE..............................................

AmendmentNo.JFJ,127 l

XV HADDAM NECK l

'?

r APR 2 6199D INDEX l,

I AU411:1STRATiVE CONTR3!S f.All SECTION I

i 6.0 ADMINISTRATIVE CONTROLS r

r 6-1 1

6.1 RESPONSIBIt1TY..........................................

s t

6.2 ORGANIZATION 6-1 l

6.2.1 ONSITE AND OFFSITE ORGANIZATIONS........................

6-1.

l 6.2.2 FACILITY STAFF..........................................

6-3 TABLE 6.2-1 MINIMUM SHIFT CREW COMPOSITION.......................

6 6-4 6.3 FACILITY STAFF 00ALIFICAT10NS...........................

i 6-5 l

6.4 TRAINING................................................

i 1

6.5 REVIEW AAP AUDIT i

6.5.1 PLANT OPERATIONS REVIEW C0KMITTEE (P0RC)................

6-5.

j i

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

6-5 l

Composition.............................................

6-6 Al t e rn at e s..............................................

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

6-6

[

Quorum..................................................

6-6

[

Responsibilities........................................

t 6-7 Authority...............................................

6-7

.l Records..................................................

NUCLEARREVIEWBOARdb#9d)

.f 6.5.2 6-7 f

Q u al i fi c at i on s..........................................

6-8 l

Composition.............................................

6;8 Consultants..............................................

6-8 Meeting Frequency.......................................

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_6 8 Quorum..................................................

6-9 Review..................................................

-i HADDAM NECK XVIII Amendment No. 125

l g

l ADMINISTRATIVE CONTROLS t

SECTION fAE Audits..................................................

6-9 l

Authority...............................................

6-10 Records.................................................

6-10 j

6.6 REPORTABLE EVENT ACTION.................................

6-11 l

6.7 SAFETY LIMIT VIOLATION..................................

6-11 6.8 PROCEDURES AND PR0 GRAMS.................................

6-11 L

i 6.9 REPORTING REOUIREMENTS 6.9.1 Routine Reports.........................................

6-13 Startup Report..........................................

6-13 6-13 l

Ann u al Re po rt s..........................................

Annual Radiological Environmental Operating Report......

6-14 f

Semiannual Radioactive Effluent Release Report..........

6-15 1

Monthly Operati ng Reports...............................

6-15 Technical Report Supporting Cycle Operation.............

6-15 l

6-17 S p e c i al Re po rt s.........................................

6.10 RECORD RETENTION.......................................

6-17 6.11 RADIATION FROTECTION PR0 GRAM............................

6-18 l

6.12 HIGH RADIATION AREA.....................................

6-19 6.13 RADIOLOGICAL EFFLUENT MONITORING AND OFFSITE DOSE

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CALCULATION MANUAL (REM 0DCM)............................

6-20 6.14 RADIO ACTIVE W ASTE TRE ATMENT.............................

6-20 3

6.15 SYSTEMS INEGRITY........................................

6-21 6[-16 3

PASS / SAMPLING AND ANALYSIS OF PLANT EFFLUENTS...........

6-21 i

b RADDAM NECK XIX Amendment ha. JJJ,155 ears t

.7uly 19, 1991-DEFINITIONS CONTAINMENT TRTEGRITY 1.6 CONTAINMENT INTEGRITY shall exist when:

All penetrations required to be closed during accident-conditions a-are either:

1)

Capable of being closed by an DPERABLE containment automatic isolation valve system, or j

2)

Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except as noted below:

Note 1)

Nomally-closed isolation valves SS-50V-150A, SS-50V-150B, SS-50V-1500, SS-50V-150D, SS-50V-151A, SS-50V-151B, SS-50V-151C, and SS-50V-151D which fail closed on loss of power and are capable of being closed within 60 seconds of a containment isolation actuation signal (CIAS) by an operator utilizing normal control switches and normal position indication within the main control room may be opened for periodic testing.

Note 2)

Nomally-closed manual isolation valves SI-V-863A. B, C, and D SA-V-413 NG 1-,ii, and SS-V-999A may be opened for periodic surveillance and containment boundary (vent and drain) manual valves may be i

opened for diagnostic _ checks to ensure Technical Specification limits or-to ensure system operability are maintained. While these valves are open, a locally stationed operator will be in direct connunication with the main control room.

This ensures the valves are capable of being close within 60 seconds of a CIAS.

t b.

The equipment hatch is closed and sealed,

)

c.

The air lock is in compliance with the requirements of Specification 3.6.1.3, i

d.

The containment leakage rates ere within the' limits of Specification 3.6.1.2, and 1

e.

The sealing mechanism associated with each penetration (e.g.,

~

i welds, bellows, or 0-rings) is OPERABLE.

CONTROLLED LEAKAGE 1.7 CONTROLLED LEAKAGE shall be that seal water flow returned from the i

reactor coolant pump number 2 seals.

]

k HADDAM NECK 1-2 AmendmentNo.Jff,,138 I

In 1 E* v s Zl I- '-'

w, ms.

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APR 2 61990 EQFER DISTRIBUTION LIMITS SURVEILLANCE Pr0VIREMENTS (Continued)

/

Measured values of core power peaking factors used in 4.2.2.2.2 determining LHGRs shall include the following allowances:

a.

Normal power peaking * **,

Flux peaking augmentation factors (Power Spike)*,' Figure 4.2-1 b.

Measurement uncertainty of 1.05, c.

d.

Statistical density factor of 1.012, Engineering factor of 1.02, e.

f.

Stack shortening / thermal expansion factor of 1.007, and Power level uncertainty of 1.02.

9 I

Items a. and b. are chosen at a core height to maximize the product.

Determined in accordance with Specification 4.2.2.2.1, using the thimble location which yields the higher total core peaking factor.

HADDAM NECK 3/4 2-7 Amendment No. 125

A 261990 -

TABLE 3.3-l.

h-REACTOR TRIP SYSTEM INSTRUMENTATION se MINIMUM TOTAL NO.

CilANNELS CllANNELS APPLICABLE OF CllANNELS TO TRIP OPERABLE MODES ACTION p

FUNCTIONAL UNIT 1.

Manual Reactor Trip 2

1 2

1, 2 1

2 1

2 3*,4*,5*

10 2.

Power Range, Neutron Flux, 4

2 3

1,2,3*,4*,5*

2, 10 Overpower Trip 3.

Wide Range, Neutron Flux, 4

2 3

2,3*,4*,5*

2, 4 liigh Start Up Rate Trip 1(a) 6f 4.

Pressurizer Pressure-Variable, low 4

2 3

5.

Pressurizer Pressure--ifiqb 3

2 2

1, 2 6f w

6.

Pressurizer Water level--liigh 3

2 2

1, 2, 3**

6#

s^

w 1(b) 6#

7.

Reactor Coolant Flow - Low a.

Above P-8 3/ loop 2/ loop 2/ loop in each in any in each operating operating operating loop loop loop IC) 6#

p b.

Above P-7 and 3/ loop 2/ loop 2/ loop I

Below P-8 in each in any two in each g

operating loops +*

operating g

loop loop g

e O.

N, u

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2 6 1990 TABLE 3.3-1 (Continuedt REACTOR TRIP _ SYSTEM INSTRUMENTATION a

MINIMUM 3E TOTAL NO.

CilANNELS CilANNELS APPLICABLE y

FUNCTIONAL UNil 0F CilANNELS TO TRIP OPERABLE MODES ACTION z

8.

Steam Flow-liigh 4 (1/ steam line) 2 1/ steam line 1, 2 9#

9.

Steam Generator Water 1/SG level 1/SG level 1/SGlevel 1, 2 57 Level-tow and coincident and with Co'ncident With Steam /Feedwater Flow 1/ steam / feed-1/ steam / feed-1/ steam / feed-Mismatch water flow water flow water flow mismatch in mismatch mismatch each SG in same loop in each SG

10. Undervoltage - Reactor 2 (1/ bus) 1 2 (1/ bus) 1(a) 8 Coolant Pumps

)

[

11. Safety Injection 2

1 2

1, 2 g (2-Ea O.a 5

W M

APR 2 61990 t

TABLE 3.3 (Continued) t TAntE NOTATION With the Reactor Trip System breakers in the c4 sed position and the j

Control Rod Drive System capable of rod withdrawal.

The low flow channel associated with trip functions derived from the out-of-service reactor coolant loop shall be in the tripped condition.

^ ^^ y.U. U,.;y,.;t.. T. y J.r ".;5. L. ;'. ; h. U ; g

y. j1:......J U.

j ov u.v.

now y.

u.

,g..........y....

i ay be bypassed when the reactor is at least 1.51rak subtritical.

The provisions of Specification 3.0.4 are not applicable.

(a) THERMAL POWER RATED THERMAL POWER.,

g Jo (b) THERMAL POWER of RATED THERMAL POWER.

j*

(c) THERMAL POWER bove 10% but elo 74% of RATED THERMAL POWER.

IC53 h

460 3rn TION STATE t4 5 ACTION 1:

With the number ri OPERABLE channels one less than the Minimum Channels j

OPERABLE requircht, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

I ACTION 2:

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

Channels, STARTUP and/or POWER OPERATION may proceed provided the following'

_l conditions are satisfied:

1 i

The inoperable channel is placed in the tripped condition within 6 a.

hours, l

The Minimum Channels OPERABLE requirement is met; however, the-b.

inoperable 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 of other. channels per Specification 4.3.1.1.

3 i

ACTION 3:

i 1

With less than the Minimum Number of Channels OPERABLE, within I hour-e a.

determine by observation of the associated permissive annunciator l

' window (s) that the interlock is in its required state for the existing j

plant condition or apply Specification 3.0.3.

l With turbine first stage pressure inoperable, continued power operation l

b.

1 may proceed provided the permissive is placed in the more conservative state for existing plant conditions.

t

'HADDAM NECK 3/4'3-5 Amendment No. 125

~

APR 2 61990

_ABLE 3.3-1 (Continued)

T ACTION STATEMENTS (Continued)

ACTION 10:

With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement for Modes 3, 4, S, 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 open the Reactor Trip System breakers within the next hour.

ACTION 11:

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 <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

(1 ACTIONJEA:

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

i HADDAM NECK 3/4 3-7 Amendment No. 125

8 i

4 J nunty 7,1992 TABLE 4.3-1 (Continued)

T_ABLE NOT ATIONS With the Reactor' Trip System breakers in the closed position and the Control Rod Drive System capable of rod withdrawal.

May be bypassed when the reactor is at least 1.5%Ak suberitiem1.T4

~

10% of RATED THERMAL POWER. Kreck & # T (a) THERMAL POWER bov w

(b) THERMAL POWE

% of RATED THERMAL POWFR.

'O T:::'"/. imn Em ;;~. L : L '._ T C.T.."! 0 T:::TT/1. ;;;T,.~

^U:S iQ 7;;:TJJ, 702:T, LL 0~.T T.".!!" T'::"."fi ^

r (1) If not performed in previous 31 days.

i above 15% of Comparison of calorimetric to excore power indication RATED THERMAL POWER.

Adjust excore channel gains consistent with

+

(2) calorimetric power if absolute difference is greater than 2%.

The are not applicable for entry into provisions of Specification 4.0.4.

MODES 1 or 2.

This requirement is not applicable when the Power Range Channels have had their gains skewed to maintain the 9% trip margin for steady state When this exception is used, a heat balance calculation will continue to be performed on a daily basis to determine core power, and conditions.

the power range channels will be verified daily to be 9% below the selected overpower trip setpoint.

Neutron detectors may be excluded from CHANNEL CALIBRATION.

(3)

The ' TRIP ACTUATING DEVICE OPERATIONAL T (4)

Reactor Trip System breakers.

-i sequent Following a refueling outage, the calibration is perfonned sThe p (5) to the plant reaching RTP.

not applicable.

\\..

?

)

1 Amendment No. TTF.147 f

HADDAM NECK 3/4 3-12 son

July 3, 1991 TABLE 3.3-2 (Continued)

TABLE NOTATIONS

• Trip function may be bypassed in this MODE when RCS pressure is ?ess than 1800 psig.

• • The channel (s) associated witn the protective functions derived from the out-of-service reactor coolant loop shall be placed in the tripped mode.

"O (a) THERMAL POWER h 07. of RATED THERMAL POW (b)

For Surveillance Testing, at most only one train may be taken out of service at a time.

(c) When feedwater control is in automatic mode.

(d)

For surveillance testing purposes, (items 3.a and 6.a of Table 4.3Y2) the minimum channels OPERABLE may be less than those specified in Table 3.3-2 for items 3.a.1, 3.a.2, and 6.a.

ACTION STATEMENTS ACTION 20 -

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 <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 />.

ACTION 21 -

With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, startup and/or power operation may proceed until performance of the next required ANALOG CHANNEL OPERATIONAL TEST provided the inoperable channel is placed in the tripped condition within I hour.

ACTION 22 -

With a channel associated with an operating loop inoperable, restore the inoperable channel to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-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 at least 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 />.

ACTION 23 -

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 et least HOT i

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 />.

HADDAM NECK 3/4 3-17 Amendment No. JEE,141

U4/40/y4-TABLE 3.12 (Continued)

ACTION STATEMENTS (Continued) l With the number of OPERABLE chtnacis one less I

Number of Channels, STARTUP and/or POW ACTION 24 The inoperable channal is placed in the tripped condition within a.

8 1 bour, and The Minimum Channels 0PERABLE requirement is met; however, one additional channel say be bypassed for up to 2 h b.

had in:. '~e than the Minimum Number of Channele associated OSCO~

ACTION 25 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> w u n' ~ by observ^"+ the interlock is in its pemissive annunciatp:

p, lant conctum., :- anly te existing p With the number of OPERABLE channels one ACTION 26 channel to OPERABLE status within j

following 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, f

With the number of OPERABLE channels one less than minimum channels OPERABLE requirement, restore the inopera ACTION 27 -

channel to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or place the DC Otherwise, be in HOT powered hydraulic pump in service. STANDBY within the 1

the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

I i

i l

i t

i i

t i

\\

l Amendment No. 7#,149 3/4 3-18 l

HADDAM NECK essa

- PR 2 61990 c '

TABLE 3.3-4 RADIATION MONITORING INSTRUNENTATION FOR PLANT OPERATIONS g-o MINIMUN E

CHANNELS CHANNELS APPLICABLE ALARM / TRIP z

SETPOINT_

ACTION y

FUNCTIONAL UNII TO TRIP / ALARM OPERABLE MODES 1.

Containment a.

RCS Leakage Detection 1)

Gaseous Radio-N.A.

I 1, 2, 3, 4 N.A.

30 activity (R-12)

ACTION STATEMENT'1 Must satisfy the ACTION requirement for Specification 3.4.6.1.

5 ACTION 30 -

Y%

I 4

m F.

?.

w i

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

a~.. -...

APR 2 61990 TABLE 3.3-8 [Ccee4eevgd).

FIRE DETECTION SiSTEMS f

Minimum Number Minimum Number Smoke Detectors Heat Detectors OPEPABLE/ Detectors OPERABLE / Detectors Location Available Available 4/4

17. Turbine building mezzanine under i

generator (T-IF) i 6/6 IB. Turbine building cranewell deluge (T-1C) i

19. Switchgear Room (New Switchgear Building) 13/13

20. Battery Room (New Switchgear Building) 2/2 i

i P

r I

h t

f.

i i

i HADDAM NECK 3/43-43 Amendment No. 125

Jcnutry 17, 1992 JEACTOR COOLANT SYSYEH 150 LATED LOOP EfMITING CONDITION FOR OPERATION

_3.4.1.5 The RCS loop stop valves of an isolated loop

• shall be shut a oither: #

The power removed from the valve operators, or a.

The boron concentration of the isolated loop shall be saintained reater than or equal to the boron concentration of the operating b.

oops.

pg APPLICABillTY: M3 DES 2

~

J ACTION:

With the requirements of the above specification not satisfied, either:

Remove power from the valve operators within one hour, or a.

Increase the boren concentration of the isolated loop to within b.

the limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or Be in at least HDT 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 C c.

SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVElttANCE RE00iREMENTS At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, if required, verify that power is 4.4.3.5.3 removed from the valve operators.

At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, if required, verify that the boron concentration of an isolated loop is greater than or equal to the boren 4.4.1.5.2 concentration of the operating loops.

A loop is considered to be isolated when the hot and cold leg stop valves are both closed.

Three-loop operation is not allowed for Cycle 17.

I i

Amendment No. J2E,148 3/4 4-9 HADDAM NECK

January 17, 1992 REACTOR COctART SYSTEM IDLEO LCOE LIMITING CONDITION FOR OPERATION The cold leg loop stop valve of an idled loop

• shall be shut and 3.4.1.8 either: i The power removed from the valve operator, or a.

concentration of the idled loop shall be saintained The boron greater than or equal to the boron concentration of the operating b.

loops.

APPLitaBIllTY: MODES

, and 2.

ACTION:

With the requirements of the above specification not satisfied, either:

Remove power from the valve operator within one hour, a.

Increase the boron concentration of the idled loop to within the b.

limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-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 c.

SWTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILtANCE REOUTREMENTS At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, if required, verify that power is 4.4.1.8.1 removed from the valve operator.

At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, if required, verify that the boron 4.4.1.8.2 concentration of an idled loop is greater that or equal to the boron concentration of the operating loops.

t A loop is considered to be idled when the hot leg stop valve is open and the cold leg stop valve is closed.

Three-loop operation is not allowed for Cycle 17.

Amendment No. J2E.146 HADDAM NECK 3/4 4-12

APR 2 61990 REACTOR COOLANT SYSTEM IDLED LOOP LIMITING CONDITION FOR OPERATION 3.4.1.9 The cold leg stop valve of an idled loop shall be shut and either:

The power removed from the valve operator, or a.

b.

T..e boron concentration of the idled loop shall be maintained greater than or equhl to the boron concentration required to meet the SHUTDOWN MARGIN requirements of Specification 3.1.1.2 or 3.1.1.3 or the refueling boron concentration of Specification

3.9.1. APPLICABILITY

MODES 3, 4, a3d6 ACTION:

With the requirements of the above specification not satisfied, either:

Remove power from the valve operator within one hour, a.

b.

Increase the boron concentration of the idled loop to within the limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or Be in at least HOT SHUTDOWN 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 c.

SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SURVEILLANCE RE001REMENTS 4.4.1.9.1 At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, if required, verify that power is removed from the valve operators.

4.4.1.9.2 At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, if required, verify that the boron concentration of an idled loop is greater than or equal to the boron concentration required to meet the SHUTDOWN KARGIN requirements of Specification 3.1.1.2 or 3.1.1.3 or the refueling boron concentration of Specification 3.9.1.

l A loop is considered to be idled when the hot leg stop valve is open i

and the cold leg stop valve is closed.

]

HADDAM NECK 3/4 4-13 Amendment No. 125 i

)

APR 2 61990

~

REACTOR COOLANT SYSTEM IDLED LOOP STARTUP LIMITING CONDITION FOR OPERATION A reactor coolant loop shall remain idled until:

3.4.1.11 The temperature at the cold leg of the idled loop is within 20*F of the highest cold leg temperature of the operating loop (s),*

a.

The boron concentration of the idled loops is greater than or b.

equal to the boron concentration required to meet the SHUTDOWN MARGIN requirements of Specification 3.1.1.2 or 3.1.1.3 or the refueling boron concentration of Specification 3.9.1.

APPLICABILITY:

MODES 3, 4 Sa. 6 j

ACTION:

With the requirements of the above specification not satisfied, do not open the idled loop cold leg stop valve.

SURVEILLANCE The idled loop cold leg temperature shall be determined to be 4.4.1.11.1 within 20*F of the highest cold leg temperature of the operating loop (s) 30 minutes prior to opening the idled loop cold leg stop valve.

within Within 30 minutes prior to opening the idled loop cold leg stop 4.4.1.11.2 valve, the idled loop shall be determined to have a boron concentration greater than or equal to the boron concentration required to meet the SHUTDOWN MARGIN requirements of Specification 3.1.1.2 or 3.1.1.3 or the If an idled loop is refueling boron concentration of Specification 3.9.1.

being started within 30 minutes after a reactor trip, this surveillance requirement may be waived if the cold leg loop stop valve is closed for less than 15 minutes.

At least once per refueling outage the stop valve / temperature 4.4.1.11.3 interlock shall be determined operable by verifying that the cold leg stop valve does not open if the cold leg temperature in the loop is more than

,20*F cooler than the highest temperature of the remaining operating loops.

At least once per refueling outage the reactor coolant pump, 4.4.1.11.4 loop stop and bypass valve interlock operability shall be demonstrated.

An operating loop (s) may be a Reactor Coolant loop (s) or a Residual Heat Removal loop (s).

HADDAM HECK 3/4 4-15 Amendment No. 125

4 9

3)

Prior to returning the valve to service following maintenance, repair, or replacement work on the valve, and 4)

Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following valve actuation due to flow through the valve.

The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4 for Specification h, above.

1.

In addition to surveillance requirement 4.4.6.2.1.g, at least once per refueling outage, perform an operational leak rate test for those portions of the HPSI, charging and RHR systems outside of containment used for or pressurized during recirculation (with the exception of RHR suction piping). The test shall be conducted at a hydrostatic pressure corresponding to the operating pressure under accident conditions.

The following provides the alternate testing for the RHR suction piping:

1.

Containment Sump to RH-MOV-22/RH-V-B R 3

'TM wor"Au.y Scae out.Eb,

Test for leakage during ILRT.

J 2.

RH-MOV-22 to PJi-CV-783 and RH-V-808A to RH-CV-808A -

Piping to be tested at a pressure of approximately 6 psi. The leak rate will be extrapolated to the operating pressure under accident conditions.

3.

Piping Downstream of RH-CV-783 and RH-CV-808A -

Piping to be tested at approximately 30 psi.

The leak rate will be extrapolated to the operating pressure under accident conditions.

"Except for those portions of the HPSI, Charging and RHR suction piping which are not testable at accident pressure during normal operation, as cefined below.

System - Those portions of HPSI suction piping downstream of the HPSI F"? on valves (SI-MOV-854A and B) and RHR/HPSI Crosstie valves (SI.MOV-gDI 902) and upstream of the HPSI pump suctions.

C m GING SYSTEM - Those portions of charging suction piping downstree. of the RHR/ Charging Crosstie Valves (RH-MOV-33A and B) and upstream of the charging pump suctions.

RHR SYSTEM - Those portions of the RHR suction piping between the containment sump and the IHR pump suctions.

The above piping will be tested in accordance with Specification 4.0.5.

HADDAM NECK 3/4 4-32a Amendment No. JJE 130

SURVEfttANCE REOUTREMENTS (Centinued)

O b.

If any periodic lype A test fails to meet 0.75 La, the test schedule for subsequent Type A tests shall be reviewed and approved

(

by the Consnission.

If two consecutive Type A tests fail to meet 0.75 La, a Type A test shall be performed at least every 18 months until two consecutive Type A tests meet 0.75 La at which time the above test schedule may be resumed or a corrective action plan may be prepared and submitted to the NRC'that provides an acceptable alternative contingent on NRC approval, c.

The accuracy of each Type A test shall be verified using the relationship:

(Lg + L,- 0.25 L,) 5 L s (LTM + l + 0.25 L,)

e o

where:

L is the percent measured containment leakage per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at TH pressure P '

t L,

is the percent superimposed leakage,.

L is the percent leakage obtained from the supplemental c

test result, and for reduced pressure tests.

L, is replaced with Lt Type B arpt ests shall be conducted at intervals no greater than l

d.

and at a pressure not less than Pa, 39.6 psig, using 24 month @ detection, soap bubble, pressure decay, or other methods halogen g.

of equivalent sensitivity, except for tests involving:

1)

Air locks, and 2)

Purge supply and exhaust isolation valves with resilient material seals.

Air locks shall be tested and demonstrated OPERABLE by the e.

requirements of Specifiestion 4.6.1.3; f.

Purge supply and exhaust isolation valves with resilient material seals shall be tested and demonstrated OPERABLE by the requirements of Specification 4.9.9; The provisions of Specification 4.0.2 are not applicable for g.

Specifications 4.6.1.2.a through 4.6.1.2.d.

T r. '. 1..;.... L.o

n.....d ;-

o in. 24.un en.n..".

iv. Tyy. ;.nj i

26 msn

n... ;, i. ;; e.l.,.

H ODAM NECK 3/4 6-3 Amendment No. JJ), 143

" ELECTRICAL POWER SYSTEMS 189M LIMITING CONDITION FOR OPERATION 3.8.1.2 As a minimum, the following A. C. electrical power sources shall i

be OPERABLE:

a.

One circuit between the offsite transmission network and the Onsite Class IE Distribution System, and b.

One diesel generator iated with the OPERABLE Onsite Class IE I

Distribution circuit, with:

1)

An engine-mounted fuel oil day tank containing a minimum volume of 400 gallons of fuel (except during engine operation),

i 2)

An underground fuel oil storage tank containing a minimum volume of 3,250 gallons of fuel,and 3)

A fuel transfer pump.

APPLICABILITY: MODES 5 and 6.

ACTES:

a.

With less than the above minimum required A.C. electrical power sources OPERABLE, immediately suspend all operations involving CORE ALTERATIONS, positive reactivity changes, movement of irradiated fuel, or crane operation with loads over the fuel storage pool. -In addition, when in MODE 5 with less than two (2) steam generators OPERABLE, or in MODE 6 with the water level less than 23 feet above the reactor vessel flange, immediately initiate corrective action to restore the required sources to OPERABLE status as soon as possible.

b.

Entry into Mode 5 pursuant to Specification 3.0.4 with less than the minimum required A.C. electrical power sources OPERABLE is not permitted.

SURVElltANEE RE0VIREMENTS 4.8.1.2 The above required A.C. electrical power sources shall be demonstrated OPERABLE by the performance of each of the requirements of Specifications 4.8.1.1.1, 4.8.1.1.2 (except for Specification 4.8.1.1.2a.5, 4.8.1.1.2b,4.8.1.1.2f),and4.8.1.1.3.

cle 16 Refueling Outage only, OPERABILITY of the Diesel Anneaeter~

would inclu e rvice water cooling via temgrehoses for a maximum of one 14 consecutive imum service water inlet of 23 psig.

temperature of 80*F and ervice water

.a The two s pumps which are powered by the "A" EDG mus

~

e construction period.

HADDAM HECK 3/4 8-7 Amendment No. J2ff,14 5 ces

APR 2 61990 4

~

V TABLE 4.8.2%

BATTERY SURVEILLANCE REOUIPEMENTS Weekly (I)

Quarterly (2) l3)

PARAMETER LIMITS FOR EACH LIMITS FOR EACH ALLOWABLE i

DESIGNATED PILOT CONNECTED CELL VALUE FOR EACH CONNECTED CELL l

CELL Electrolyte

> Minimum level

> Minimum level Above top of Level indication mark, indication mark,

plates, and < %" above and < 4' above and not maximum level maximum level overflowing j

indication mark indication mark Float Voltage 12.10 volts 12.10 volts 12.07 volts Specifi{4) 11.200(5) 11.190 Not more than 0.020 below the Gravity i

average of all connected cells average of all Average of all connected cells connec{ggcells

>1.200

>1.195 TABLE NOTATIONS II)

For any Weekly parameter (s) outside the limit (s) shown, the battery may be considered OPEPABLE provided that within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> all the Quarterly measurements are taken and found to be within their allowable values, and provided all Weekly and Quarterly parameter (s) are restored to within limits within the next 6 days.

(2)

For any Quarterly parameter (s) outside the limit (s) shown, the battay may be considered OPERABLE provided that the Quarterly parameters are-within their allowable values and provided the Quarterly parameter (s) are restored to within limits within 7 days.

I3) Any Quarterly parameter not within its allowable.value indicates an inoperable battery.

I4) Corrected for electrolyte tem; hture and level.

(5) Or battery charging current is less than 2 amps when on charge.

l HADDAM NECK 3/4 8-10 Amendment No. 125

January 7.'1992 REFUELING OPERATIONS 3/4 A.

INSTRUMENTATION l

A\\

NG CONDITION FOR OPERATION i

As a.uinimum, two Source Range Neutron Flux Monitors shall be and o>erating, each with continuous visual indication in the i

3.9.2.a control room anc one with audible indication in the c OPERABLE room when CORE ALTERATIONS or positive re Flux Monitor shall be OPERABLE and with a visual indication in the control room and audible at least one Source Range Neutron operating indication in the containment.

Source Range High Neutron Level Alaras l'

be OPERABLE and operating with a minimum As a minimum, two 3.9.2.b logic to audibly alam in both the control room and containment shall (Cont:inMint Evacuation) of two (2).

APPLICABillTYt MODE 6.

ACTION:

With one of the above required monitors inope a.

positive reactivity changes.

With both of the required monitors inoperable or not operating.

• determine the boron concentration of the Reactor Coolant System b.

least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

SURVEltLANCE RE001REMENTS 4.9.2 Each Source Range Neutron Flux Monitor shall be demonstra i

OPERABLE by perfomance of:

A CHANNEL CHECK at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, a.

An ANALOG CHANNEL OPERATIONAL TEST within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> A

l b.

initial start of CORE ALTERATIONS, and An ANALOG CHANNEL OPERATIONAL TEST at least once per 7 d; c.

r Amendment No. //f,147 3/4 9-2 H.ADDAM NECK ns

IU;R 2 6;1990 REFUELING OPERATIONS 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE [POOOBVILDING v

LIMITING CONDITION FOR OPERATION Loads in excess of 1650 pounds shall be prohibited from travel over 3.9.7

' fuel assemblies in the-storage pool.

APPLICABILITY: With fuel assemblies in the storage pool.

ACTION:

With the requirements of the above specification not satisfied, a.

place the crane load in a safe condition.

b.. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REOUIREMENTS Administrative controls that prevent the travel of loads in excess 4.9.7 of 1650 pounds over fuel assemblies shall be in place prior to lifting a load in excess of 1650 pounds.

f HADDAM NECK 3/4 9-7 Amendment No. 125

January 17,_1992

t REACTIVITY CONTROL SYST mS

}/4.1

-t BASES-z 3/4.1.1 B0 RATION CONTROL ggg 3/4.1.1.1. 3/4.1.1.2. e4 3/4.1.1.3 $ HUT 00WW MARGIN l

(1) the reactor can be made A sufficient SHUTDOWN MRGIN ensures that:

i nts-suberttical from all operating conditions, (2) the reactivity trans e l

ithin

.l associated with postulated accident conditions are controllab e w i tly acceptable limits, and (3) the reactor will be sa dition.

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f fuel 3

SHUTDOWN MARGIN requirements vary throughout core life as The most restrictive l

depletion. RCS boron concentration, and RCS T f-cycle life (EOL), and is l

conditionduringMODES1,2and3occursateWo.

li RCS associated with a postulated steam line break accident and resu t n for four loop operation and 2600 pcm for three d

cocidown.

d d by f

Operation in MODE 3 with two operating reactor coolant pumps the four loop steam line break analysis. Operation in MODE 3 with l

loops (both-j operating reactor coolant pump and two OPERABLE reactor coola loop stop valves open in each loop)'is bounded by the three l

Because of the short time involved, the 2600 pcm SHUTDOl l

MARGIN limit need not be applied to the closure of the cold leg s l

break analysis.

in order to restart the reactor coolant pumps from' an initia l

In the analysis of this j

operation condition. associated with the boron dilution accident.

i ed accident, a minimum SHUTDOWN MARGIN of 3100 pcm in MOD j

GIN to control the reactivity transient. Accordingly, t]

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current design basis assumptions.

i 5 MODERATOR TmPERATURE COEFFICIENT 3 /4.1.1.(

d to The limits on the moderator temperature coeff ccMition assumed in the accident and transient analysis.

i fication are associated with a specific set of ditions other than those l

4 MTC values of this speci a conditions; measurement of MTC values at con explicitly stated with extrapolation to the specified co e

acceptable.

temperature and boron concentration.

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Amendment No. JJ),148 l

B3/41-1 MADDAM NECK l

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January 17. 1992 i

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REACTIVITY CONTR0t SYS1 Ql1 BASES si6DERATOR TEMPERATURE COEFFICIENT (Continued) f The Surveillance Require'eent for measurement of the MTC at the b within its limits the fuel cycle is adequate to confirm that the MTC rar-I RCS boron concentration associated with fuel burnup.

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. 3/4.1.1.s MINIMUM TEMPERATURE FOR CRITICALITY i

This specification ensures that the reactor will not be made critical wit This the Reactor Coolant System average temperature less than 525'F.

is within it analyted temperature range, (2) the trip instrumentation is' limitation is required to ensure:

within its normal operating range, (3) the pressurizer is capable of being.

in an CPERABLE status with a steam bubble, and (4) the reactor vessel i temperature.

above its minimum RTNDT 3/4.1.2 BORATION SYSTEMS The boration systems ensure that negative reactivity control is availablel during each MODE of facility operation.(1) borated water sources, (2) charging p Tracing Systems, any(6) an emergency power suppl this function include:

l generators.

With the RCS average temperature above 200'F a minimum of two boron injection flow paths are required to ensure single functional capability The the event an assumed failure renders one of the flow paths inoperable.

l boration capability of either flow path is sufficient to l

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after xenon decay and cooldown to 200*F.

capability requirement occurs at EOL from full power eq j

3 borated water from the boric acid tank meets this requirement.

i With the RCS temperature below 200*F, one boration system is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restriction system becomes inoperable.

Amendment No. JJE 148L B3/41-2 HADDAN NECK

Janu::ry 17. 1992 3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section provide assurance of fuel integr'ity during Condition I (Normal Operation) and !! (Incidents of Mo equal to I.30 during norsal operation and in short-ters transients, and (2) cy) events by:

liatting the fission gas release, fuel pellet temperature, and cladding In addition, mechanical properties to within assumed design criteria.

liciting the peak linear power density during Condition I eve and the ECCS Interim Acceptance Criterion limit of 2300*F peak cladding temperature for stainless steel clad fuel and the 10CFRSO.46 and Appendix K licit of 2200*F peak cladding temperature for zircaloy fuel are not cxceeded.

3/4.2.1 AXIAL OFFSET The AXIAL OFFSET specification provides continuous confirr,ation of accept-able LINEAR HEAT GENERATION kAILS (LHGR) during the time interval between incore measurements.

3/4.2.2 LINEAR HEAT GENERATION RATE Limiting the peak LINEAR HEAT GENEPATION RATE (LHGR) during Condition I cvents provides assurance that the initial condition assumed for LOCA cnalyses are met and the peak cladding temperature limits are not exceeded.

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NUCLEAR ENTHALPY RISE HOT CHANNEL FACTO 3/4.2.3 The limit on the NUCLEAR ENTHALPY RISE HOT CHMNEL FA the minimum DNBR limit is not exceeded.

is measurable, but will normally only be determined periodically as N

The F This periodic speciNed in Specification 4.2.3.1.2 and 4.2.3.2.2.

surveillance is sufficient to insure that the limits are maintained provided:

The control rod insertion limits provided in the TECHNICAL REPORT a.

SUPPORTING CYCLE OPERATION are maintained, and The AIIAL OFFSET limits provided in the TECHNICAL REPORT SUPPORTING b.

CYCLE OPERATION are saintained.

The relaxation of F" fur all permissible rod insertion limits.as a functio The full power radial power shape limits include a 4% incere measurement uncertainty.

3/4.2.4 00ADRANT POWER TitT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power distribu-

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Radial tion satisfies the design values used in power capability analysis.

Amendment No. J2E,ue HADDAM NECK B3/4 2-1 essa

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INSTRUMENTATION BASES 3.4.3.3.6 FIRE DETECTION INSTRUMENTATION (Continued) equipment and is an integral element in the overall facility Fire Protection i

Program.

Fire detectors that are used to actuate Fire Suppression Systems represent a more critically important component of a plant's Fire Protection Program than detectors that are installed solely for early fire war cation.

greater.

The loss of detection capability for Fire Suppression S As a result, the establishment of a fire watch patrol must be initiated at an earlier stage than would be warranted for the loss of any area.

detectors that provide only early fire warning. The establishment of fre-quent fire patrols in the affected areas is required to RADIOACTIVE L10VID EFFt0ENT MONITORING INSTRUMENTA 3/4.3.3.7 The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid The effluents during actual or potential releases of liquid effluents.

Alarm / Trip Setpoints for these instruments shall be the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20.

The OPERABILITY and use of this instrumentation is consistent requirements of General Design Criteria 60, 63, and 64 of Appendix A to

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10 CFR Part 50, RADI0 ACTIVE GASEOUS _EFFt0ENT MONITORING INSTRU 3/4.3.3.8 The radioactive gaseous effluent instrumentation is provided to monitor an control, as applicable, the releases of radioactive materials in gaseous The effluents during actual or potential releases of gaseous in accordance with the methodology and parameters in the REMODCM to ensure that the alarq/ trip will occur prior to exceeding the limits ofThe n

10 CFR Part 20.

Ja1 Design Criteria 60, 63, and 64 of tent with the requirements of Ge; Appendix A to 10 CFR Part 50.

3/4.3.3.9 BORON DILUTION ALARM The shutdown monitors provide indication of positive reactivity insertion during operation in Modes 3, 4, 5, and 6.

Boron Dilution design basis analysis.

T ern w n 3/4.3.4 AFLOOD M PROTECTION The liquid level instrumentation is provided to monitor liquid levels in The system areas of potential flooding caused by local pipe ruptures.

l ensures that early warning will occur so that protective action can be take Amendment No. 125 B3/4 3-3 HADDAM HECK

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1 INSTRUMENTATION 1

1 BASES ZuTEt.WAL FLodb VitorEcrwh3 3/4.3.4-TLOOOM;0 Af tontinued)

U in the event of a. localized flooding condition in areas of the plant that i

The loss of detection capability represents house safety-related equipment.

As a result the a degradation of flooding protection for any area.

I establishment of a liquid level watch patrol must be initiated at an early

.l The establishment of frequent liquid level watch patrols in the

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affected areas is required to provide detection capability until the inoper-stage.

able instrumentation is restored to OPERABILITY.

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HADDAM NECK B3/4 3-4 Amendment No. 125 I

I APR 2 61990 REACTOR COOLANT SYSTEM BASy 3

4 R CTOR COOLANT SYSTEM LOOPS AND COOLANT CIRCULATION (Continued) ictions on starting an RCP with one or more RCS cold legs less than The or equal to 315*F are provided to prevent RCS pressure transients, caused by energy additions from the Secondary Coolant System, which could exceed the limits of Appendix G to 10 CFR Part 50. The RCS will be protected against

+

overpressure transients and will not exceed the limits of Appendix G by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 20*F above each of the RCS cold leg temperatures.

The requirement to maintain the boron concentration of an isolated / idled loop greater than or equal to the baron concentration of the operating loops or the boron concentration required to meet SHUTDOWN MARGIN requirements ensures that no unacceptable reactivity addition to the core could occur during startup of an isolated / idled loop. Verification of the boron concen-tration in an isolated / idled loop prior to opening the stop valves provides a reassurance of the adequacy of the boron concentration in the isolated /fdled loop.

Startup of an isolated / idled loop could inject cool water from the loop into the core. The reactivity transient resulting from this cool water injection is minimized by prohibiting isolated / idled loop startup until its tempera-ture is within 20*F of the operating loops.

3/4.4.2 SAFETY VALVES The pressurizer Code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2735 psig. The required relieving capacity of each safety valve is 240,000 lbs. per hour at 2,485 psig as assumed in the safety analysis. Each safety valve is conservatively designed to relieve 293,300 lbs. per hour of saturated steam at 2485 psig.

The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown.

In the event that no safety valves are OPERABLE, an operating RHR loop, connected to the RCS, provides overpressure relief capabili;y and will prevent RCS overpressur-ization.

In addition, the Overpressure Protection System provides a diverse means of protection against RCS overpressurization at low temperatures.

During operation, all pressurizer Code safety valves must be OPERABLE to prevent the RCS from being pressurized above its Safety Limit of 2735 psig.

The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from a complete loss-of-load assuming no P4 actor trip until the first Reactor Trip System Trip Setpoint is reached (i.e., no credit is taken fcr a direct Reactor trip on the loss-of-load) and also assuming no operation of the power-operated relief valves or steam dump valves.

Demonstration of the safety valves' lift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASHE Boiler and Pressure Code.

HADDAM NECK B3/4 4-2 Amendment No. 125

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APR 2 6 ISBU CONTAINMENT SYSTEMS fasES t

3 /4. 6.1. 5 AIR TEMPERATURE (Continued) and a main steam line break inside the containment. Measurements shall be taken from all OPERABLE temperature detectors to determine the average air temperature.

3/4.6.1.6 CONTAINM T STRUCTURAL INTEGRITY l

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This limitation ensures that the structural integrity of the containment will be maintained comparable to the original design standards for the life Structural integrity is required to ensure that the of the facility.

containment will withstand the maximum pressure of 39.6 psig in the event of a LOCA. A visual inspection in conjunction with the Type A leakage tests is sufficient to demonstrate this capability.

3/4.6.1.7 CONTAINMENT VENTILATION SYSTEM l

The 42-inch containment purge supply and exhaust isolation valves and the 8-inch bypass valve are required to be closed and locked closed during plant operation since these valves have not been demonstrated capable of closing during a LOCA or steam line break accident. Maintaining these valves locked closed during plant operations ensures that excessive quantities of radio-active materials will not be released via the Containment Purge System. To provide assurance that these containment valves cannot be inadvertently opened, the valves are locked closed in accordance with Standard Review Plan 6.2.4 which includes mechanical devices to lock the valve closed.

I Containment post accident hydrogen venting can be accomplished by two methods. One uses the containment air particulate monitoring system and the other uses the containment purge exhaust system. These methods are not required in any short time frame after an accident; it is expected that months may elapse.

In any event, if the systems are not operable because of maintenance reasons, they can be made operable.

System operability can be readily obtained provided access into the containment is not required.

Containment purge is utilized as a back-up means of venting hydrogen from the containment following a loss-of-coolant i..cident. The containment air rarticulate monitoring system provides the primary means of purging because provides adequate purge flow to prevent an explosive mixture build-up 41e allowing fine control of the release of radioactivity during purges.

2n necessary to effect repairs'to the containment purge or purge bypass isolation valves, a blank flange must be applied to the 42" purge air exhaust penetration inside the reactor containment so that the containment remains leak tight. This renders the purge system inoperable for a finite time. Seven days is considered a reasonable length of time for repair parts to be received, installed and the system retested for leak tightness and returned to service.

i HADDAM NECK B3/4 6-2 Amendment No. 125

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' 1/d.7 PLANI mitM i

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. RASES 1/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY VALVES b

The OPERABILITY of the main steam line Code safety valves ensures that thel 1105,.(1100 secondary Coolant System pressure will be limited to below psia), of its design pressure of 1000 psia during the most seversThe maximum reliev anticipated system operational transient.

cssociated with a Turbine trip from 2005 RATED THERMAL' POWER coincide cn assumed loss of condenser heat sink (i.e., no steam bypass to the condenser).

The specified valve lift settings and relieving capacities are in accordance l

with the requirements of Section XI of the ASME Boiler and Pre Code, 1971 Edition.

lbs/hr.which is 120% of the total j

all of the steam lines is g,504,000lbs/hr at 100% RATED THERMAL POWER.

secondary steam flow of 7,872,000 i

3 /4.7.1.7 AUXILI ARY FEEDWATER SYSTEM i

The OPERABILITY of the Auxiliary Feedwater System ensures that the Reactor Coolant System can be cooled down to less than 350'F from normal operating l

conditions in the event of a total loss of offsite power.

Each steam tu,rbine-driven auxiliary feedwater pump has a capacity sufficient t

to ensure adequate delivery of feedwater flow to remove With'one auxiliary feedwater Residual Heat Removal System operating range.

1 pump inoperable, the safest mode of operation is HOT SHUTDOWN with the With of being provided by the RHR System.

1 heat removal function capable two steam turbine-driven feedwater pumps inoperable, at l

declared inoperable, or be in HOT STANDBY within the next six~ hours an l

HOT SHUTDOWN with the following six hours.be restored to OPERAB first pump or be in HOT STANDBY in the next six hours and HGT SHUTDOWN l

within the following six hours.

.l The auxiliary feedwater (AFW) system's design basis requires AFW to be i

automatically initiated and to be independent of any AC electrical powl source for at least two hours.

1 hydraulic pump, controls, and DC power supply are required to be OPERA l

for the associated ATW pump to be OPERABLE.

l instrumentation does not function, the associated AFW pump remains OPERABLl as long as the DC powered hydraulic pump is started and maintained operatingj in accordance with the stated ACTION statement.

3/4.7.1.3 AUXILIARY FEEDWATER SUPPLY l

The OPERABILITY of the demineralized water storage tank (DWST) and primary with the minimum water volume ensures that i

water storage tank (PWST sufficient ~ water is avai able' to maintain the RCS at HOT STANDl for 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> with steam discharge to the atmosphere concurrent with total The contained water volume limit includes an' loss-of-offsite power.

allowance for water not usable because of tank discharge line location or other physical characteristics.

149 B3/4 7-1 Amendment No. J/J.A g HADDAM NECK asse x-1

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i I N SESrr._.1, In In addition, the auxiliary feedwater system can be initiated manually.

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this case, feedwater is available from the DWST by gravity feed Within is adequate for decay heat removal for a period of a auxiliary feedwater pump.

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PLANT SYSTEMS l

E.A. 5fl.

AUXILI ARY FEEDWATER SUPPLY (Continued) 150 gpm. Makeup water is available during this period from the PWST which contains a minimum volume of 80,000 gallons. The PWST transfer pumps can transfer 200 gpm from the PWST to the DWST. An alternate supply can be provided from the 100,000 gallons Recycled Primary Water Storage Tank.

3/4.7.1.4 SPECIFIC ACTIVITY i

The limitations on Secondary Coolant System specific activity ensure that the resultant offsite radiation dose will be limited to a small fraction of 10 CFR Part 100 dose guideline values in the event of a steam line rupture.

This dose also includes the effects of a coincident 0.4 gpm reactor-to-secondary tube leak in the steam generator of the affected steam line.

These values are consistent with the assumptions used in the safety analyses.

3/4.7.1.5 MAIN STEAM LINE TRIP VALVES The OPERABILITY of the main steam line trip valves ensures that no more than one steam generator will blowdown in the event of a steam line rupture.

This restriction is required to: (1) minimize the positive reactivity effects of the Reactor Coolant System cooldown associated with the blowdown, and (2) limit the pressure rise within containment in the event the steam line rupture occurs within containment. The OPERABILITY of the main steam line trip valves within the closure times of the Surveillance Requirements are consistent with the assumptions used in the safety analyses.

3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE tiMITATION The limitation on steam generator pressure and temperature ensures that t'e pressure-induced stresses in the steam generators do not exceed the ma>:aem allowable fracture toughness stress limits. The limitations of 70*F anc 200 psig are based on a steam generator RTNDT of 10*F and are sufficient to prevent brittle fracture. The heatup and cooldown rate of 100*F/hr for the steam generators are specified to ensure that stresses in these vessels are maintained within acceptable design limits.

?>'.7.3 SERVICE WATER SYSTEM t

4 The OPERABILITY of the Service Water 'ystem ensures that sufficient cooling capacity is available for continued operation of safety-related equipment f.: t[q g,.] ); T,p..ai:n, 0.t.,3 pu{ingnj{pa{aggagcjpen},congi}{ons.,

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: ::: ;' :: ;;i;. The two service water pumps which are powered by the "A" EDG must be operable during the construction period. The redundant cooling capacity of this system, assuming a single failure, is consistent

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with the assumptions used in the safety analysis. A service water header is comprised of the two service water pumps associated with each diesel generator and the HADDAM NECK B3/4 7-2 Amendment No. JJ1I,145 sess

APR 2 61990 PLANT SYSTEMS

_ BASES 3/4.7.9 FEEDWATER ISOLATION VALVES The accident analysis for a main steam line break assu Also, the closure of these valves based on a CIAS is credited in determining the Pressure / Temperature limits for the p signal (CIAS).

i environmental qualification.to the feedwater regulation valves in the event a falls open during a Main Steam Line Break.

3/4.7.10 EXTERNAL FLOOD PROTECTION The thresholds regarding flood protection ensure that fa The estimated Connecticut River be made) in the event of flood conditions. probable maximum flo Normal flood control measures water level), is 39.5 feet mean sea level.

provide protection to safety-related equipment to El. 30 feet mean sea Normal f?ood protection to this elevation is based on a lowBased on the level.

probability of exceedance and structural capacity limitations.

one to two day rise period of the PMF, alternative means of providing decay hot removal for flooding events up to the PMF is provided in A0P 3.2-24.

CLEANUP SYSTEM 3/4.7.11 PRIMARY AUXIt1 ARY BUILDING PAB Air Cleanup System consists of two exhaust fans, two prefilters, a HEPA-HECA filter assembly, and interconnecting ductwork.

Air cleanup is accomplisited using one exhaust fan, one prefilter, the HEPA-HECA filter, and interconnected ductwork.

The radiological consequences analyses for loss-of-coolant accidents assume Primary Auxiliary Building efficiencies which are ensured by this Technica Also, in consideration of a fuel handling accident inside Specification.

containment, (i.e., when the containment is being purged) the purge discharge would be directed through the Primary Auxiliary Building charc Credit is again taken for these filters in reducing the filters.

radiological consequences.

Amendment No. 125 B3/4 7-6 HADDAM NECK

APR 2 61991) 3/4.

1 ELECTRICAL POWER SYSTEMS v

BASES 3/4.8.1. 3/4.8.2 AND 3/4.

A. C. SOURCES. D. C. SOURCES. DNSITE POWER DISTRIBUTION The OPERABILITY of the A. C. and D. C. power sources and associated distribution systems during operation ensures that sufficient power will be available to supply the safety-related equipment required for: (1) the safe shutdown of the facility, and (2) the mitigation and control of accident The minimum specified independent and conditions within the facility.

redundant A.C. and D.C. power sources and distribution systems satisfy the requirements of General Design Criterion 17 of Appendix A to 10 CFR Part 50.

The ACTION requirements specified for the levels of degradation of the power sources provide restriction upon continued facility operation comensurate The OPERABILITY of the power,ources is with the level of degradation.

consistent with the initial condition assumptions of the safety analyses and based upon maintaining at least one redundant set of onsite A.C. and D.C.

power sources and associated distribution systems OPERABLE during accident conditions coincident with an assumed loss-of-offsite power and single When one diesel generator is failure of the other onsite A.C. source.

r inoperable, there is an additional ACTION requirement to verify that the charging pump, HPSI pump, LPSI pump and RHR pump that depend on the remaining OPERABLE diesel generator as a source of emergency power, are also This requirement is intended to provide assurance that a OPERABLE.

loss-of-offsite power event will not result in a complete loss of safety function of critical systems during the period one of the diesel generators is inoperable. The term, verify, as used in this context means to administratively check by examining logs or other information to determine if certain components are out-of-service for maintenance or other reasons.

It does not mean to perform the Surveillance Requirements needed to demonstrate the OPERABILITY of the component.

The OPEPABILITY of the minimum specified A.C. and D.C. power sources and associated distribution systems durirg, shutdown and refueling condition ensures that:

(1) the facility car, be maintained in the refueling or shutdown condition for extended time periods, and (2) sufficient

' instrumentation and control capability is available for monitoring and maintaining the facility status.

The Surveillance Requirements for demonstrating the OPERABILITY of the diesel generators are based on the recomendations of Regulatory Guides 1.9,

" Selection of Diesel Generator Set Capacity for Standby Power Supplies",

March 10, 1971; 1.108, " Periodic Testing of Diesel Generator' Units Used as Onsite Electric Power Systems at Nuclear Power Plants", Revision 1, August 1977; and 1.137, " Fuel-0il Systems for Standby Diesel Generators", Revision 1, October 1979, and guidance given in Generic Letter 84-15.

h HADDAM NECK 43/4 8-1 Amendment No.125

APR 2 6199D 4

REFUELING OPERATIONS BASES 3/4.9.6 MANIPUL ATOR CRANE The OPERABILITY requirements for the manipulator cranes ensure that:

(1) manipulator cranes will be used for movement of control rod drive shafts and fuel assemblies, (2) each crane has sufficient load capacity to lift a drive shaft or fuel assembly, and (3) the core internals and reactor vessel are protected from excessive lifting forces in the event they are inadvertently ged during lifting operations.

3/4a9.

CRANE TRAVEL - SPENT FUEL STORAGE BUILDING (Ar striction on movement of loads in excess of the nominal weight of a fue and control rod assembly and associated handling tool over other fuel assemblies in the storage pool ensures that in the event this load is (1) the activity release will be limited to that contained in dropped:

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 analysis.

3/4.9,8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCUUi10N The requirement that at least one RHR LOOP be in operation ensures that: (1) sufficient cooling capacity is available go remove decay heat and maintain the water in the reactor vessel below 140 F as required during the REFUELING MODE, 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 23 feet of water above the reactor vessel flange, a large heat sink is available for core cooling. Thus, in the event of a failure of the operating RHR LOOP, adequate time is provided to initiate emergency procedures to cool the core.

3/4.9.9 CONTAINMENT PURGE SUPPLY. PURGE EXHAUST. AND PURGE EXHAUST BYPASS

9tATION SYSTEM OPERABILITY of this system ensures that the containment vent and purge

,aetrations can be isolated upon detection of high radiation levels within the containment. The OPEPABILITY of this system is required to restrict the release of radioactive material from the containment atmosphere to the environment.

HADDAM HECK B3/4 9-2 Amendment No. 125

t July 9, 1990 3/4.10 SPECI AL TEST EXCEPT10NS BASES 3/4.10.1 SHUTDOWN MAPGIN a minimum amount of control rod This Special Test Exception provides that immediately available for reactivity control when tests are per-worth is formed for control rod worth measurement.

This special test exception is required to permit the periodic verification of the actual versus predicted core reactivity condition occurring as a result of fuel burnup or fuel cycling operations.

3/4.10.2 PHYSICS TESTS This Special Test Exception permits PHY3ICS TESTS to be performed at less slightly lower than or equal to 5% of RATED THERMAL POWER with the Rt.5 T than normally allowed so that the fundamental nuticar char $Ueristics of the core and related instrumentation can be verified.

In order for various characteristics to be accurately measured, it is at times necessary. to restrictions of these Technical Specifications.

operate outside the normalFor instmce, to measure the moderator temperature necessar.~ to position the various control rods at heights which may not neemally be allowed by Specification 3.1.3.6 which in turn may cause the RCS T

to fall slightly below the minimum temperature of Specification 3 p3,4, a

3/4.10.3 POSITION INDICATION SYSTEM-SHUTDOWN This Special Ted Sception permits the Position Indication Systems to be inoperable during rod drop time measurements.

The exception is required since the data necessary to determine the rod drop time are derived from the This l

induced voltage in the position indicator coils as the rod is dropped.

induced voltage is small compared to the normal voltage and, therefore, cannot be observed if the Position Indication Systems remain OPERABLE.

3/4.10.4

. " [ POSITION INDICATION SYSTEM - OPERATING i

This Special Test Exception permits the IRPI system to be inoperable during performance of data collection / verification / adjustment testing of the The testing is required to develop and implement correction factors the individual rod position indicator.

While the IRPI system is IRPl.

the _ indicated individual rod position cannot be used to verify for each ir. operable, or control rod insertion control rod alignment (Specification 3.1.3.1) limits (Specifications 3.1.3.5, 3.1.3.6.1 and 3.1.3.6.2).

The actual rod position for banks C, D and A is, however, unaffected by this testing.

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HADDAM NECK B3/4 10-1 Amendment No.

, 127 i

i

fD'n % o bdU 3/4.11 RADIOACTIVE EFFLUENTS PASES 3/4.11.1 L10VID EFFLUENTS 3/4.11.1.1 CONCENTRATION This specification is provided to ensure that the concentra be less than the concentration levels specified in 10 CFR Part 20, Appendix This limitation provides additional assurance that B, Table II, Column 2.the levels of radioactive materials in bodies of water outside result in exposures within:

(1) the Section II.A design objectives of Appendix I,10 CFR Part 50, to a MEMBER OF THE PUBLIC, and (2) the limi The concentration limit for dissolved 10 CFR 20.106(e) to the population.

or entrained noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent contentration in water using the methods describ F

3/4.11.1.2 005E LIGot D S This specification is provided to implement the requirements of SectionsThe Limiting C II.A, III.A, and IV.A of Appendix I, 10 CFR Part 50.

for Operation implements the guides set forth in Section II.A of Appendix 1.

The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A The dose calculation methodology kept "as low as is reasonably achievable".

and parameters in the REMODCM implement the requirements in Section III.A Appendix I that conformance with the guides of Appendix I is to be shown by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate path to be substantially underestimated.

for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in Regulatory Guide I.109,

• Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I,* Revision 1, October 1977, and Regulatory Guide 1.113, " Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I,* April 1977.

Amendment No. 125 B3/4 11-1 HADDAM NECK

.~

l 6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILI]1 Mg i

bi 9t 1Mo.t - dc00'd*

The 5g 5 gtet* Di-+etee-shall be responsible for overall l

6.1.1 facility operation and shall delegate, in writing, the succession to this responsibility during his absence.

i 6.2 ORGANIZATION 6.2.1 ONSITE AND OFFSITE ORGAN 17ATIONS Onsite and offsite organizations shall be established for unit operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting the safety of the nuclear power plant.

l l

Lines of aethority, responsibility, and comunication shall be a.

established and defined for the highest management levels through intermediate levels to and including all operating organization positions, These relationships shall be documented and updated, as appropriate, in the f0rm of erganization Charts, functional t

descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements shall be documented in the Quality Assurance Topical Report.

i V;ee. Pr* ick:1~- k ocus MecM The Nudur-?y"'-

--te shall be responsible for overall unit l

l b.

safe operation and shall have control over those onsite activities i

necessary for safe operation and maintenance of the plant.

The Executive Vice President-Nuclear, shall have corporate l

c.

responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to ti1..

plant to ensure nuclear safety.

l d.

The individuals who train the operating staff and those who carry i

out health physics and quality assurance functions may report to l

i the appropriate onsite manager; however, they shall have-sufficient organizational freedom to ensure their independence from operatit.g pressures.

l 6.2.2 FACILITY STAFF Each on-duty shift shall be composed of at least the minimum shift a.

~

crew composition shown in Table 6.2-1; b.

At least one licensed Operator shall be in the control room when fuel is in the reactor.

In addition, while the facility is in PODE 1, 2, 3 or 4, at least one itcensed Senior Operator shall be in the control room; i

HADDAM NECK 6-1 Amend ient No. JJE,155 sc c i

l j

o,.

I ADMINISTRATIVE CORTROLS f.

Successful completion of the Thames Valley State Technical College associate's degree in Nuclear Engineering Technology program, provided that the individual was enrolled in the program by October 1, 1987.

2.

Dedicated STA: Must meet the STA training criteria of NUREG-0737, Item I.A.I.1, and have received specific training in plant design, and response and analysis of the plant for transients and accidents.

6.4 TRAINING A retraining and replacement training program for the facility staff 6.4.1 shall be maintained under the direction of the Nuclear Unit Director and l

shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and 10CFR55.59. The Director-Nuclear Training has the overall responsibility for the implementation of the Training Program.

A training program for the Fire Brigade shall be maintained under 6.4.2 the direction of the Director-Nuclear Training and shall meet or exceed the l

Intent of Section 27 of the NFPA Code-1975, except for Fire Brigade training sessions which shall be held at least quarterly, 6.5 REVIEW AND AUDIT 6.5.1 PLANT OPERATIONS REVIEW COMMITTEE (PORC)

D)NCTION 6.5.1.1 The PORC shall function to advise the h:1 car St:ticmC4Wctee on l

all matters related to nuclear safety.

COMPOSITION 6.5.1.2 The PORC shall be composed of the:

$s:cle~ T - 400 t % ^leck viat fr

' " " " %eter--

Chairperson-Member:

Nuclear Unit Director Member:

Operations Manager Member:

Maintenance Manager Member:

Instrument and Control Manager Member:

Reactor Engineer Member:

Engineering Manager Member:

Nuclear Services Director Member:

Plant Quality Services Supervisor Member:

Chemistry Manager Member:

Health Physics' Manager Member:

Security Manager 3

HADDAM NECK 6-5 Amendment No. JJJ, JJJ,155 sere

09/25/92 ADMINISTRATIVE CONTROLS ALTERNATES All alternate a mbers shall be appointed in writing by the PORC 6.5.1.3 Chairperson to serve on a temporary basis; however, no more than two alternates shall participate as voting members in PORC activities at any one time.

MEETING FREOUENCY The PORC shall meet at least once per calendar month and as 6.5.1.4 convened by the PORC Chairperson or his/her designated alternate.

000 RUM 6.5.1.5 The quorum of the PORC shall consist of the Chairperson or his/her

~

designated alternate and four members including alternates.

RESPON IBILITIES 6.5.1.6 The PORC shall be responsible for:

Review of: (1) all procedures required by Specification 6.8 and a.

changes thereto, and 2) any other proposed procedures or changes l

thcrete as determined by the gie r 5iauon uiredur to affect nuclear safety; vg e N, & 4 _ % p s c.

b.

Review of all proposed tests and experiments that affect nuclear safety; Review of all proposed changes to the Technical Specifications; c.

d.

Review of all proposed changes or modifications to plant systems or equipment that affect nuclear safety; Investigation of all violations of the Technical Specifications, e.

including the preparation and forwarding of reports covering evaluation and recommendations to prevent recurrence, to the Executive Vice President-Huclear and to the Chairperson-of the Nuclear Review Board; f.

Review of all REPORTABLE EVENTS; Review of facility operations to detect potential safety hazards; g.

Performance of special reviews, investigations or analyses and h.

reports' thereon as requested by the Chairpersra of the Nuclear l

Narrlear-4t4 tier. Director; Review Board or the \\

Wrc % sded-Ilararw 9<l' HADDAM NECK 6-6 Amendment No. Jyy,155 ooro

0 i

ADMINISTRATIVE CONTROLS l

1.

Review of the Security Plan and implementing procedures and shall submit recommended changes to the Chairperson of the Nuclear Review Board; I

j.

Review of the Emergency Plan and implementing procedures.and shall submit recommended changes to the Chairperson of the Nuclear Review Board; AUTHORITY

' b" M cM 6.5.1.7 The PORC shall:

d Report to and be advisory to the-Nue;;;r StWer. Dimte on those l

a.

areas of responsibility specified in Section 6.5.1.6(a) through l

(j);

b.

Render deterzinations in writing to the "=ltar-St: tier Ofr;;ter if any item considered under Specification 6.5.1.6a. through d.,

above, as appropriate and as provided by 10CFR50.59 or 10CFR50.92 constitutes an unreviewed safety question or requires a significant hazards consideration determination.

Provide written notification, meeting minutes may be used for this c.

purpose, to the Executive Vice President-Nuclear and the Chairperson of the Nuclear Review Board of disagreement betwee the PORC and the A ; leer........

.. % nowever, the Nwc4eer Statter Direeter shall have responsibility for resolution of such disagreements pursuant to Specification 6.1.1.

RECORDS 6.5.1.8 The PORC shall maintain written minutes of each meeting that, at a minimum, document the results of all PORC activities performed under the responsibility and authority provisions of these Technical Specifications.

A Cop 4fs shall be provided to theteuu i "'- '- "- ' ""-'-

g/l

\\

Chairperson of the Nuclear Review Board.

6.5.2 NUCLEAR REVIEW BOARD (NRB)

'IFICATIONS

' 2.1 The minimum qualifications of NRB members are as follows:

a.

The Chairperson and NRB members shall have:

l 1.

an academic degree in engineering or physical science field, or hold a senior management position, and 2.

a minimum of five years technical experience in their respective field of expertise, and 3.

a minimum of nine (9) years combined academic and technical j

experience.

i HADDAM NECK 6-7 Amendment No. JJJ,155 l'

sere i

r

ADMINISTRATIVE CDNTROLS i

t.

The ;ikB shall collectively have the experience and competence required to review activities in the following areas:

t 1.

Nuclear power plant operations 2.

Nuclear engineering 3.

Chemistry and radiochemistry 4.

Metallurgy 5.

Nondestructive testing 6.

Instrumentation and control 7.

Radiological safety 8.

Mechanical and electrical engineering 9.

Administration 10.

Quality assurance practices l/IEfre-SN - NA00cn Idtt.k COMPOSITION The NRB shall consist of no less than eight,!nor more than eleven 6.5.2.2 members including the Chairperson and the Mhr Lt;;th: 04-rter. The Chairperson and members of the NRB shall be appointed in writing by the Executive Vice President - Nuclear.

l CONSULTANTS 3

6.5.2.3 Consultants shall be utilized as determined by the NRB Chairperson l

to provide expert advice to the NRB.

MEETING FREOUENCY 6.5.2.4 The NRB shall meet at least once per 6 months.

DUORUM i

6.5.2.5 The quorum of the NRB necessary for the performance of the NRB review and audit functions of these Technical Specifications shall consist of at least enough members to constitute a majority-of the assigned members

~ including the Chairperson or a designated alternate. No more than a l

minority of the quorum shall have line responsibility for operation of the facility.

i i

i i

HADDAM NECK 6-8 Amendment No. 125,155 CC7C

[

i

i i

i

\\ ADMINISTRATIVE CONTROLS r

The applicable procedures recommended in Appendix A of Regulatory a.

Guide 1.33, Revision 2, February 1978;-

I b.

The requirements and recommendations of Sections 5.1 and 5.3 of ANSI N 18.7-1976.

j Fire Protection Program im' plementation.

i c.

d.

Quality controls for effluent monitoring, using the guidance in W Regulatory Guide 1.21 Rev.1, June 1974.

l t

RADIOLOGICAL EFFLUENT MONITORING AND OFFSITE DOSE CALCULATION e.

MANUAL (REMODCM) implementation except for Section I.E.

Radiological Environmental Monitoring.

l f.

PROCESS CONTROL PROGRAM implementation.

W.ct kesM. b #A004 / deck G

l 6.8.2 Each procedure of Specification 6.8.1, and changes-thereto, shall be \\

reviewed by the PORC and shall be approved by the L.; lear St tier Oirect:n V l

prior to implementation and reviewed periodically as set forth in each

[

~

document or in administrative procedures.

6.8.3 Temporary changes to procedures of Specification 6.8.1 may be made provided:

b The intent of the original procedure is not altered; a.

i b.

The change is approved by two members of the plant management staff, at least one of whom holds a Senior Operator license on the 4

unit affected; and j

The change is documented, reviewed by the PORC and approved by the c.

L ;i nt Stctie. 01 !He within 14 days of implementation.

l Vice Presioe~s-or>w a< ek.

6.8.4 Written procedures I all be established, implemented and maintained i

covering Section 1.E., Radiological Environmental Monitoring, of the l

REMODCM.

6.8.5 All procedures and procedure changes required for the Radiological-l Environmental Monitoring Program of Specification 6.8.4 above 'shall be

- reviewed by an individual (other than the author) from the Radiological-Assessment Branch or the Production Operation Services Laboratory (POSL).and-i approved by appropriate supervision.

1

. Temporary changes may be made provided the. intent of the original procedure i

is not altered and the change is documented and reviewed by an individual

-l (other than the author) from the Radiological Assessment Branch or the POSL, o

within 14 days of implementation.

HADDAM NECK 6-12 Amendment No. JJJ,155 ante

-l

.~.

6.13 RADIOLOGICAL EFFLUENT MONITORING AND OFFSITE DOSE

_ CALCULATION MANUAL (REMODCM)

Section I, Radiological Ef ficents Mr.,nitoring Manual, shall outline the samplino and analysis programs to determine the concentration of radioactive materiais released offsite as well as dose comitments to individuals in those exposure pathways and for those radionuclides released as a result of facility operation.

It shall also specify operating guidelines for RADI0 ACTIVE WASTE TREATMENT SYSTEMS and report content.

Section II, the Offsite Dose Calculation Manual, shall describe the methodology and parameters to be used in the calculation of offsite doses due to radioactive gaseous and liquid effluents and in the calculations of gaseous and liquid effluent monitoring instrumentation Alarm / Trip Setpoints consistent with the applicable LCO's contained in these Technical Specifications.

Changes to the REMODCH:

Shall be documented and records of reviews perfomed shall be retained a.

as required by Specification 6.10.3.m.

This documentation shall contain:

1)

Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change (s), and 2)

A determination that the change will maintain the level of 1

radioactive effluent control required by 10 CFR 20.106, 40 CFR Part 193, 10 CFR 50.36a, and Appendix I to 10CFR Part 50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations.

b.

Shall become effective after review and acceptance by PORC and the l

ni -" w approval of the """-

"-N Wet Precit~T - 4 ADaw d<ck Shall be submitted to the Comission in the form of a complete, lecible c.

copy of the entire REMM or ODCM, as appropriate, as a part of er concurrent with the Semiannual Radioactive Effluent Release Report for the period of the report in which any change was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g., month / year) the change was implemented.

6.14 RADI0 ACTIVE WASTE TREATMENT Procedures for liquid and gaseous Hioactive effluent discharges from the facility shall be prepared, approvec, maintained and adhered to for all operations involving offsite releases of radioactive effluents. These procedures shall specify the use of appropriate RADIDACTIVE WASTE TREATMENT SYSTEMS utilizing the guidance provided in the REMODCM.

The Solid RADIDACTIVE WASTE TREATMENT SYSTEM shall be operated in accordance with the PROCESS CONTROL PROGRAM to process wet radioactive wastes to meet shipping and burial ground requirements.

MADDAM NECK 6-20 Amendment No. JJJ,155 acto

6.15 SYSTEMS INTEGRITY t

The licensee shall implement a program to reduce leakage from systems r

cutside containment that would or could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. This program shall include the,following:

Provisions establishthg preventive maintenance and periodic visual a.-

inspection requirements, and b.

Integrated leak test requirements for each system at a frequency not to exceed refueling cycle intervals.

CAPSf 6.16 PASS /Samnlino and Analysis of Plant Effluents The licensee shall implement and maintain a program which will ensure the 1

capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant gascous effluents, and containment atmosphere samples under accident conditions. This program shall include the following:

a.

Training of personnel b.

Procedures for sampling and analysis, and Provisions for maintenance of sampling and analysis equipment.

c.

t i

i

?

l I

1 I

i 1

1 HADDAM NECK 6-21 Amendment No. JJJ.155 sore I

Docket No'. 50-213 814369 l

k i

t Haddam Neck Plant Proposed Revision to Technical Specifications Editorial Cleanup of Technical Specifications i

i 5

5 i

i I

i i

.1 t

i March 1993

~

l i

d m

r

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0VIREMENTS SECTION PAGE 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems...............................

3/4 4-29 Ope rati onal Le akage.....................................

3/4 4-31 3/4.4.7 CHEMISTRY...............................................

3/4 4-33 TABLE 3.4-1 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS..............

3/4 4-34 TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS............................

3/4 4-35 3/4.4.8 SPECIFIC ACTIVITY.......................................

3/4 4-36 g

FIGURE 3.4-2 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY GREATER THAN 1 microcurie / gram DOSE EQUIVALENT I-131........

3/4 4-37 TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM.................................

3/4 4-38 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Re acto r Cool ant Sys t em..................................

3/4 4-39 i

FIGURE 3.4-3 CONNECTICUT YANKEE LIMIT CURVE FOR HYDROSTATIC AND LEAK TESTING APPLICABLE FOR 22.0 EFFECTIVE FULL l

POWER YEARS.................

3/4 4-41 FIGURE 3.4-4 CONNECTICUT YANKEE REACTOR COOLANT SYSTEM HEATUP l

LIMITATIONS FOR 22.0 EFFECTIVE FULL POWER YEARS.....

3/4 4-42 FIGURE 3.4-5 CONNECTICUT YANKEE REACTOR COOLANT SYSTEM C00LDOWN l

LIMITATIONS FOR 22.0 EFFECTIVE FULL POWER YEARS.....

3/4 4-43 Pressurizer..........................................

3/4 4-45 low Temperature Overpressure Protection Systems......

3/4 4-46 3/4.4.10 STRUCTURAL INTEGRITY....................................

3/4 4-48.

3/4.4.11 REACTOR COOLANT SYSTEM VENTS............................

3/4 4-49 l

3/4.4.12 FAILED FUEL R0DS........................................

3/4 4-SI.

l

.{

i HADDAM NECK VIII Amendment No. //E, J/E 0104

'I

INDEX LIMITING CONDITIONS-FOR OPERATION AND SURVEllLANCE RE0VIREMENTS SECTION PAGE 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ECCS SUBSYSTEMS - Tavg GREATER THAN OR EQUAL TO 350*F...

3/4 5-1 l

TABLE 4.5-1 SAFETY INJECTION ACTUATED AUTOMATIC VALVES...........

3/4 5-6 l

TABLE 4.5-2 ECCS MANUAL VALVES...................................

3/4 5-6 3/4.5.2 ECCS SUBSYSTEMS - Tavg LESS THAN 350'F..................

3/4 5-7 3/4.5.3 REFUELING WATER STORAGE TANK............................

3/4 5-9 j

3/4.5.4 pH CONTROL SYSTEM.......................................

3/4 5-10 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT j

Containment Integrity...................................

3/4 6-1 i

Containment Leakage.....................................

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

3/4 6-4 I nt e rn al Pre s s u re.......................................

3/4 6-6

(

Air Temperature.........................................

3/4 6-7 Containment Vessel Structural Integrity.................

3/4 6-8 Containment Ventil ati on System..........................

3/4 6-9 3/4.6.2 CONTAINMENT AIR RECIRCULATION SYSTEM....................

3/4 6-10 3/4.6.3 CONTAINMENT ISOLATION VALVES............................

3/4 6-12 1

3/4.7 PLANTS SYSTEMS 3/4.7.1 TURBINE CYCLE Safety Valves...........................................

3/4 7-1 j

TABLE 3.7-1 STEAM LINE SAFETY VALVES PER L00P....................

3/4 7-2

)

Auxiliary Feedwater System..............................

3/4 7-3 Auxil i ary Feedwater Supply..............................

3/4 7-4 Specific Activity.......................................

3/4 7-5 i

1 HADDAM NECK IX Amendment No. J2E, 0104

INDEX r

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0UIREMENTS l

i SECTION PAGE t

t TABLE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM..........................

3/4 7-6 Main Steam Line Trip Va1ves.............................

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

3/4 7-8 l

3/4.7.3 SERVICE WATER SYSTEM....................................

3/4 7-9 3/4.7.4 SNUBBERS................................................

3/4 7-10 j

TABLE 4.7-2 SNUBBER VISUAL INSPECTION INTERVAL...................

3/4 7-10a 3/4.7.5 SEALED SOURCE CONTAMINATION.............................

3/4 7-14 3/4.7.6 FIRE SUPPRESSION SYSTEMS Fire Water Supply / Distribution System..............

3/4 7-16 i

Spray and/or Sprinkle r Systems..........................

3/4 7-19 00 Systems.............................................

3/4 7-21 2

Halon Systems...........................................

3/4 7-22 Fire Stations...........................................

3/4 7-23 TABLE 3.7-4 FIRE STATIONS........................................

3/4 7-24 Yard Fire Hydrants and Associated Fire Hose Houses......

3/4 7-25 TABLE 3.7-5 YARD FIRE HYDRANTS...................................

3/4 7-27 3/4.7.7 FIRE RATED ASSEMBLIES...................................

3/4 7-28 3/4.7.8 FLAMMABLE LIQUIDS CONTR0L...............................

3/4 7-30 3/4.7.9 FEEDWATER ISOLATION VALVES..............................

3/4 7-31 TABLE 3.7-6 FEEDWATER ISOLATION VALVES...........................

3/4 7-32 3/4.7.10 EXTERNAL FLOOD PROTECTION...............................

3/4 7-33 3/4.7.11 PRIMARY AUXILIARY BUILDING AIR CLEANUP SYSTEM...........

3/4 7-34 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES 0perating...............................................

3/4 8-1 TABLE 4.8-1 DIESEL GENERATOR TEST SCHEDULE.......................

3/4 8-6 Shutdown................................................

3/4 8-7 3/4.8.2 D.C. SOURCES 0perating...............................................

3/4 8-8 TABLE 4.8-2 BATTERY SURVEILLANCE REQUIREMENTS....................

.3/4 8-10 Shutdown................................................

3/4 8-11 HADDAM NECK X

Amendment No. J/J, JJJ, 0104

fNDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0VIREMENTS SECTION PAGE 3/4.8.3 ONSITE POWER DISTRIBUTION 0perating...............................................

3/4 8-12 Shutdown................................................

3/4 8-14 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION.....................................

3/4 9-1 3/4.9.2 INSTRUMENTATION.........................................

3/4 9-2 1

3/4.9.3 DECAY TIME..............................................

3/4 9-3 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS.......................

3/4 9-4 r

3/4.9.5 COMMUNICATIONS..........................................

3/4 9-S 3/4.9.6 MANIPULATOR CRANE.......................................

3/4 9-6 i

3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE BUILDING..............

3/4 9-7 3/4.9.8 RESIDUAL HEAT REMOVAL AND C0OLANT CIRCULATION r

High Water Leve1........................................

3/4 9-8

[

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

3/4 9-9 3/4.9.9 CONTAINMENT PURGE SUPPLY, PURGE EXHAUST, AND l

PURGE EXHAUST BYPASS ISOLATION SYSTEM...................

3/4 9-10

[

3/4 9.10 WATER LEVEL - REACTOR VESSEL............................

3/4 9-11 l

3/4.9.11 WATER LEVEL-STORAGE P00L...............................

3/4 9-12 3/4.9.12 FUEL STORAGE BUILDING AIR CLEANUP SYSTEM................

3/4 9-13 5/4.10 SPECIAL TEST EXCEPTIONS

{

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

3/4 10-1 3/4.10.2 PHYSICS TESTS...........................................

3/4 10-2 3/4.10.3 POSITION INDICATION SYSTEM - SHUTD0WN...................

3/4 10-3 3/4.10.4 POSITION INDICATION SYSTEM - OPERATING..................

3/4 10-4 l

3/4.11 RADI0 ACTIVE EFFLUENTS f

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

3/4 11-1

[

Dose, Liquids...........................................

3/4 11-2 HADDAM NECK XI Amendment No. JJE, JJ/,

[

t 0104 L

-+

INDEX i

BASES SECTION FAG.E i

aff_.0 APPLICABILITY...........................................

B 3/4 0-1 f

6 t

3/4.1 REACTIVITY CONTROL SYSTEMS i

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

B 3/4 1-1 3/4.1.2 B0 RATION SYSTEMS........................................

B 3/4 1-2 3/4.1.3 MOVABLE CONTROL ASSEMBLIES..............................

B 3/4 I-3 i

1 3/4.2 POWER DISTRIBUTION LIMITS l

i 1

3/4.2.1 AXIAL 0FFSET............................................

B 3/4 2-1 3/4.2.2 LINEAR HEAT GENERATION RATE.............................

B 3/4 2-1 N

3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR F B 3/4 2-1

[

AH............

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

B 3/4 2-1

{

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

B 3/4 2-2 l

i i

3/4.3 INSTRUMENTATION t

i i

3/4.3.1 & 3/4.3.2 REACTOR TRIP SYSTEM INSTRUMENTATION AND ENGINEERED i

SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION........

B 3/4 3-1

{

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

B 3/4 3-2 3/4.3.4 INTERNAL FLOOD PROTECTION...............................

B 3/4 3-4 i

3/4.4 REACTOR COOLANT SYSTEM i

.i 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION..........

B 3/4 4-1 j

3/4.4.2 SAFETY VALVES..........................................

B_3/4 4-2 3/4.4.3 PRESSURIZER............................................

B 3/4 4-3

.+

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

B 3/4 4-3 3/4.4.5 STEAM GENERATORS.......................................

B 3/4 4-3 l

3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE......................

B 3/4 4 5 l

i

-HADDAM NECK XIII Amendment No. JJE, 0105

k t

INDEX f

f BASES SECTION PAGE t

i 3/4.4.7 CHEMISTRY..............................................

B 3/4 4-7 l

3/4.4.8 SPECIFIC ACTIVITY......................................

B 3/4 4-7 l

3/4.4.9 PRESSURE / TEMPERATURE LIMITS............................

B 3/4 4-8 3/4.4.10 STRUCTURAL INTEGRITY...................................

B 3/4 4-12 l

3/4.4.11 REACTOR COOLANT SYSTEM VENTS...........................

B 3/4 4-12 s

3/4.4.12 FAILED FUEL R0DS.......................................

B 3/4 4-13 l'

l 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 & 3/4.5.2 ECCS SUBSYSTEMS................................

B 3/4 5-1 3/4.5.3 REFUELING WATER STORAGE TANK...........................

B 3/4 5-2 3/4.5.4 pH CONTROL SYSTEM......................................

B 3/4 5-2 i

3/4.6 CONTAINMENT SYSTEMS i

l t,

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

B 3/4 6-1 3/4.6.2 CONTAINMENT AIR RECIRCULATION SYSTEM....................

B 3/4 6-3 l

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

B 3/4 6-3 1

3/4.7 PLANT SYSTEMS i

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

B 3/4 7-1

.l 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION.........

B 3/4 7-2

a 3/4.7.3 SEP,VICE WATER SYSTEM....................................

B 3/4 7-2 l

3/4.7-.4 SNUBBERS................................................

B 3/4 7-3 3/4.7.5 SEALED SOURCE CONTAMINATION.............................

B 3/4 7-4 f

3/4.7.6 FIRE SUPPRESSION SYSTEMS................................

B 3/4 7-4 f

3/4.7.7 FIRE RATED ASSEMBLIES...................................

B 3/4 7-5 t

3/4.7.8 FLAMMABLE LIQUIDS CONTR0L...............................

B 3/4 7-5 l

3/4.7.9 FEEDWATER ISOLATION VALVES..............................

B 3/4 7-6 l

3/4.7.10 EXTERNAL FLOOD PROTECTION...............................

B 3/4 7-6 3/4.7.11 PRIMARY AUXILIARY BUILDING AIR CLEANUP S;5 TEM...........

B 3/4 7-6 l

HADDAM NECK XIV Amendment No. J U,

02cs

{

j INDEX l

t BASES SECTION PAGE l

I 3/4.8 ELECTRICAL POWER SYSTEMS 5

t 3/4.8.1, 3/4.8.2 and 3/4.8.3, A.C. SOURCES, D.C.SC'JRCES, ONSITE l

POWER DISTRIBUTION........................................

B 3/4 8-1 f

3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION.....................................

B 3/4 9-1 3/4.9.2 INSTRUMENTATION.........................................

B 3/4 9-1 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 MANIPULATOR CRANE.......................................

B 3/4 9-2 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE BUILDING..............

B 3/4 9-2 t

3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION...........

B 3/4 9-2 l

3/4.9.9 CONTAINMENT PURGE SUPPLY, PURGE EXHAUST, AND PURGE EXHAUST BYPASS ISOLATION SYSTEM...................

B 3/4 9-2 3/4.9.10 & 3/4.9.11 WATER LEVEL - REACTOR VESSEL AND STORAGE P00L....................................................

B 3/4 9-3 t

3/4.9.12-FUEL STORAGE BUILDING AIR CLEANUP SYSTEM................

B 3/4 9-3 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN.........................................

B 3/4 10-1 j

3/4.10.2 PHYSICS TESTS...........................................

B 3/4 10-1 3/4.10.3 POSITION INDICATION SYSTEM - SHUTD0WN...................

B 3/4 10-1 3/4.10.4 POSITION INDICATION SYSTEM - OPERATING..................

B 3/4 10-1 i

3/4.11 RADIOACTIVE EFFLUENTS t

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

B 3/4 11-1 3/4.11.2 GASEOUS EFFLUENTS.......................................

B 3/4 11-2

.l 3/4.11.3 TOTAL 00SE..............................................

B 3/4 11-3 j

HADDAM NECK XV Amendment No. //J, J//,

0105 i

+

1

INDEX i

ADMINISTRATIVE CONTROLS i

SECTION PAGE l

6.0 ADMINISTRATIVE CONTROLS 6.I RESPONSIBILITY.........................................

6-1 i

6.2 ORGANIZATION l

6.2.1 ONSITE AND OFFSITE ORGANIZATIONS........................

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

6-1 TABLE 6.2-1 MINIMUM SHIFT CREW COMPOSITION.......................

6-3 I

6.3 FACILITY STAFF 0VALIFICATIONS...........................

6-4 54 TRAINING................................................

6-5 i

6.5 REVIEW AND AUDIT r

4 6.5.1 PLANT OPERATIONS REVIEW COMMITTEE (P0RC)................

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

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

6-5 Alternates..............................................

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

5-6 b

Quorum..................................................

6-6 I

Responsibilities........................................

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

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

6-7 6.5.2 NUCLEAR REVIEW BOARD (NRB)..............................

6-7

-l l

Qu al i fi c at i on s..........................................

6-7 j

Composition.............................................

6-8

{

Consultants.............................................

6-8 Meeting Frequency.......................................

6-8 Quorum..................................................

6-8 Review..................................................

6-9 HADDAM NECK XVIII Amendment No. JEE, 0106

)

I

1 i

l INDEX ADMINISTRATIVE CONTROLS l

SECTION PAGE l

Audits..................................................

6-9 Authority...............................................

6-10 Records.................................................

6-10 4

6.6 REPORTABLE EVENT ACT10N.................................

6-11 l

l E

6.7 SAFETY LIMIT VIOLATION..................................

6-11 6.8 PROCEDURES AND PR0 GRAMS.................................

6-11 i

6.9 REPORTING RE0VIREMENTS i

1 I

6.9.1 Routine Reports.........................................

6-13 Startup Report..........................................

6-13 An nu al Re p o r t s..........................................

6-13

.j Annual Radiological Environmental Operating Report......

6-14

{

Semiannual Radioactive Effluent Release Report..........

6-15 l

1 Monthly Operating Reports...............................

6-15 l

Technical Report Supporting Cycle Operation.............

6-15 i

t Sp e c i al Re p o rt s.........................................

5-17 l

6.10 RECORD RETENT10N........................................

6-17 1

6.11 RADIATION PROTECTION PR0 GRAM............................

6-18 4

6.12 HIGH RADIATION AREA.....................................

6-19 l

t t

~6.13 RADIOLOGICAL EFFLUENT MONITORING AND OFFS 11E DOSE i

CALCULATION MANUAL (REM 0DCM)..................

6-20 6.14-RADI0 ACTIVE WASTE TREATMENT..............................

6-20 r

i~'

6.15 SYSTEMS INEGRITY........................................

6-21 6.16 PASS / SAMPLING AND ANALYSIS OF PLANT EFFLUENTS............ 21 s

HADDAM NECK XIX Amendment No. J/E, JEE, 0106

-J

DEFINITIONS CONTAINMENT INTEGRITY 1.6 CONTAINMENT' INTEGRITY shall exist when:

a.

All penetrations required to be closed during accident conditions are either:

1)

Capable of being closed by an OPERABLE containment automatic isolation. valve system, or 2)

Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except as j

noted below:

Note 1)

Normally-closed isolation valves SS-50V-150A, SS-S0V-1500, SS-SOV-1500, SS-50V-150D, SS-S0V-151A, SS-S0V-1518, SS-50V-151C, and SS-50V-151D which fail-closed on loss of power and are capable of being closed within 60 seconds cf a containment isolation actuation signal (CIAS) by an operator utilizing normal control switches and normal position indication within the main control room may be opened for periodic testing.

Note 2)

Normally-closed manual isolation valves SI-V-863A, B, C, and D, SA-V-413, and SS-V-999A may be opened for l

l periodic surveillance and containment boundary (vent i

and drain) manual valves may be opened for diagnostic checks to ensure Technical Specification limits or to i

ensure system operability are maintained. While i

these valves are open, a locally stationed operator will be in direct communication with the main control room. This ensures the valves are capable of being closed within 60 seconds of a CIAS.

b.

The equipment hatch is closed and sealed, c.

The air lock is in compliance with the requirements of Specification 3.6.1.3, d.

The containment leakage rates are within the limits of Specification 3.6.1.2, and e.

The sealing mechanism associated with each penetration (e.g., welds, bellows, or 0-rings) is OPERABLE.

CONTROLLED LEAKAGE 1.7 CONTROLLED LEAKAGE shall be that seal water flow returned from the reactor coolant pump number 2 seals.

HADDAM NECK 1-2 Amendment No. JEE, J7E,.

0107

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) i 4.2.2.2.2 Measured values of core power peaking factors used in determining LHGRs shall include the following allowances:

a.

Normal power peaking * **,

b.

Flux peaking augmentation factors (Power Spike)*,

l l

c.

Measurement uncertainty of 1.05, d.

Statistical density factor of-1.012, i

e.

Engineering factor of 1.02, f.

Stack shortening / thermal expansion factor of 1.007, and g.

Power level uncertainty of 1.02.

l i

i i

i

~l i

Items _a. and b. are chosen at a core height to maximize the product.

Determined in accordance with Specification 4.2.2.2.1, using the thimble-location which yields-the higher total core peaking factor.

'l r

i HADDAM NECK 3/4 2-7 Amendment No. J/E, caos.

l

TABLE 3.3-1 ll REACTOR TRIP SYSTEM INSTRUMENTATION

.a E

MINIMUM m:

TOTAL NO.

CHANNELS CHANNELS APPLICABLE fj FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 1.

Manual Reactor Trip 2

1-2 1, 2 1

2 1

2 3*,4*,5*

10 2.

Power Range, Neutron Flux, 4

2 3

1,2,3*,4*,5*

2, 10 Overpower Trip 3.

Wide Range, Neutron Flux, 4

2 3

2,3*,4*,5*

2, 4 High Start Up Rate Trip 4.

Pressurizer Pressure-Variable, low 4

2 3

1(a) 6#

5.

Pressurizer Pressure--High 3

2 2

1, 2 6#

s

[

6.

Pressurizer Water Level--High 3

2 2

1, 2, 3***

6#

l 7

7.

Reactor Coolant Flow - Low a.

Above P-8 3/ loop 2/ loop 2/ loop 1(b) 6#

in each in any in each operating operating operating loop loop loop b.

Above P-7 and 3/ loop.

2/ loop 2/ loop 1(c) 6#

M Below P-8 in each in any two in each El operating loops **

operating 8

loop loop

?.

F

?

l

TABLE 3.3-1 (Continued) g REACTOR TRIP SYSTEM INSTRUMENTATION 8

3-MINIMUM TOTAL NO.

CHANNELS CHANNELS APPLICABLE 2

p; FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION n

8.

- Steam Flow-High 4 (1/ steam line) 2 1/ steam line 1, 2 9#

9.

Steam Generator Water 1/SG level 1/SGlevel 1/SG 1evel 1, 2 5#

Level-Low and coincident and Coincident With with

- Steam /Feedwater Flow 1/ steam / feed-1/ steam / feed-1/ steam / feed-Mismatch

- water flow water flow water flow mismatch in mismatch mismatch each SG in same loop in each SG 10.

Undervoltage - Reactor.

2 (1/ bus) 1 2 (1/ bus) 1(a) 8 R

Coolant Pumps a

{ 11. Safety Injection' 2

1 2

1, 2 12 i

a.

lit

- a n

6 m

v-,,.w.m,%,_.e,w.

,,,,,,<r.,.,m.,6,.

c,,,-m.,.%

,,.,, -.-,.w#w w.m., c.i ea e.w,4

,.or.

,,...ese -,. m ea, s

..-,-..,.-,..w

• .e.

,,+-ee,,

.,,.pe-.-

e

. -,,,.g e.,,--

,+,g<>

-ys-,w--

,.s,-ema

+-

JL--

,.u.i.

41 2

j TABLE 3.3-1 (Continued) i TABLE NOTATION With the Reactor Trip System breakers in the closed position and the Control Rod Drive System capable of rod withdrawal.

3

-l The low flow channel associated with trip functions derived from the out-of-service reactor coolant loop shall be in the tripped condition.

May be bypassed when the reactor is at least 1.5%Ak subtritical.

The provisions of Specification 3.0.4 are not applicable.

(a) THERMAL POWER greater than 10% of RATED THERMAL POWER.

l (b) THERMAL POWER greater than or equal to 74% of RATED THERMAL POWER.

l.

(c) THERMAL POWER greater than 10% but less than 74% of RATED THERMAL POWER.

l' ACTION STATEMENTS j

ACTION 1:

l With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status i

within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

j ACTION 2:

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

Channels, STARTUP and/or POWER OPERATION may proceed provided the.following I

conditions are satisfied:

a.

The inoperable channel is placed in the tripped condition within 6

hours, l

b.

The Minimum Channels OPERABLE requirement is met;

however, the i

inoperable 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 of other channels per Specification 4.3.1.1.

L ACTION 3:

a.

With less than the Minimum Number of Channels OPERABLE, within l' hour determine by observation of the associated permissive annunciator window (s) that the interlock is in its required state for the existing l

plant condition or apply Specification 3.0.3.

4 b.

With turbine first stage pressure inoperable, continued power operation may proceed provided the permissive is placed in the more conservative state for existing plant conditions.

HADDAM NECK 3/4 3-5 Amendment No. U E, 0110 f

TABLE 3.3-1 (Continued)

ACTION STATEMENTS (Continued)

ACTION 10:

With the number of OPERABLE channels one less than the Minimum Channels l

OPERABLE requirement for Modes 3, 4, 5, restore the inoperable channel to l

OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or open the Reactor Trip System breakers within the next hour.

l ACTION 11:

I With the number of OPERABLE channels one less than the Minimum Channels l

OPERABLE requirement, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

i ACTION 12:

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

i I

t i

i i

l

[

HADDAM NECK 3/4 3-7 Amendment No. JJE, 0111 r

I

t TABLE 4.3-1 (Continued)

TABLE NOTATIONS l

With the Reactor Trip System breakers in the closed position and the j

Control Rod Drive System capable of rod withdrawal.

i May be bypassed when the reactor is at least 1.5%Ak subcritical.

(a) THERMAL POWER greater than 10% of RATED THERMAL POWER.

(b) THERMAL POWER greater than or equal to 74% of RATED THERMAL POWER.

j i

(1)

If not performed in previous 31 days.

(2) Comparison of calorimetric to excore power indication above 15% of

[

RATED THERMAL 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 for entry into l

MODES 1 or 2.

This requirement is not applic.able when the Power Range Channels have had their gains skewed to maintain the 9% trip margin for steady state conditions. When this exception is used, a heat balance calculation will continue to be performed on a daily basis to determine core power, and i

the power range channels will be verified daily to be 9% below the l

selected overpower trip setpoint.

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

(4) The TRIP ACTUATING DEVICE OPERATIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip attachments of the Reactor Trip System breakers.

(5) Following a refueling outage, the calibration is performed subsequent to the plant reaching RTP.

The provisions of Specification 4.0.4 are not applicable.

I 1

1 t

HADDAM NECK 3/4 3-12 Amendment No. J2E, J M,

0112

t I

l TABLE 3.3-2 (Continued) i TABLE NOTATIONS

• Trip function may be bynassed in this MODE when RCS pressure is less than 1800 psig.

• • The channel (s) associated with the protective functions derived from the I

out-of-service reactor coolant loop shall be placed in the tripped mode.

(a) THERMAL POWER greater than 10% of RATED THERMAL POWER.

l 1

(b)

For Surveillance Testing, at most only one train may be taken out of service at a time.

l (c) When feedwater control is in automatic mode.

(d)

For surveillance testing purposes, (items 3.a and 6.a of Table 4.3-2) the l

minimum channels OPERABLE may be less than those specified in Table 3.3-2 for items 3.1.a, 3.a.2, and 6.a.

l ACTION STATEMENTS i

ACTION 20 With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT l

STANDBY within 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 i

following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

ACTION 2I With the number of OPERABLE channels one less than the Minimum. Channels OPERABLE requirement, startup and/or power operation may proceed until performance of the next required ANALOG CHANNEL OPERATIONAL TEST provided the inoperable channel is placed in the tripped condition within I hour.

ACTION 22 With a channel associated with an operating loop inoperable, l

restore the inoperable channel to OPERABLE status within-4 hours 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 at least 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 />.

-l ACTION 23 With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable '

channel to 0PERABLE 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 l

within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

1 HADDAM NECK 3/4 3-17 Amendment No. JJE, J,4J, 0113 l

l

~

TABLE 3.3-2 (Continued)

ACTION STATEMENTS (Continued)

With the number of OPERABLE channels one less than the Total ACTION 24 Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a.

The inoperable channel is placed in the-tripped condition within I hour, and b.

The Minimum Channels OPERABLE requirement is met; however, one-

~

additional channel may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance test'ing of other channels per Specification 4.3.2.1.

ACTION 25 -

Not used.

-l ACTION 26 -

With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel 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 reduce the THERMAL POWER to below 10% of RATED THERMAL POWER within the following I hour.

ACTION 27 -

With the number of OPERABLE channels one less than the minimum channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or place the DC powered hydraulic pump in service. Otherwise, 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 HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

t i

l i

HADDAM NECK 3/4 3-18 Amer.dment No. J/E, JfS, Olle i

--- ------ J

~

~

TABLE 3.3-4 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS E

MINIMUM

=

CHANilELS CHANNELS APPLICABLE ALARM / TRIP 2

p FUNCTIONAL UNIT TO TRIP / ALARM OPERABLE MODES SETPOINT ACTION l

1.

Containment a.

RCS Leakage Detection 1)

Gaseous Radio-N.A.

I 1, 2, 3, 4 N.A.

30 activity (R-12)

ACTION STATEMENT l

ACTION 30 -

Must satisfy the ACTION requirement for Specification 3.4.6.1.

Y F

Et ar 3a m.

. -m

_______,m______________.~-,w_,,--_,

,,..,,,,b,

,_d_.

____,_,,-_m.-.,,

2.

,,_.._.m_...m.,

n _., _,..

,_.,_..,,_,..,__,,m,.,,.,,mm.,,

,,,.,y

B TABLE 3.3-8 (Continued) l I

i FIRE DETECTION SYSTEMS

-i Minimum Number Minimum Number Smoke Detectors Heat Detectors.

OPERABLE / Detectors OPERABLE / Detectors Location Available Available

17. Turbine building mezzanine under 4/4 l

generator (T-lF) 1 6/6 18.

Turbine building cranewell deluge (T-lC)

19. Switchgear Room (New Switchgear Building) 13/13 i

i 20.

Battery Room (New Switchgear Building) 2/2 i

i l

l i

i i

HADDAM NECK 3/4 3-43 Amendment No. JJE, 0116 l

REACTOR COOLANT SYSTEM ISOLATED LOOP LIMITING CONDITION FOR OPERATION 3.4.1.5 lhe RCS loop stop valves of an isolated loop

• shall be shut and either: #

a.

The power removed from the valve operators, or b.

The baron concentration of the isolated loop shall be maintained greater than or equal to the boron concentration of the operating loops.

APPLICABILITY: MODES 1 and 2 l

ACTION:

With the requirements of the above specification not satisfied, either:

a.

Remove power from the valve operators within one hour, or b.

Increase the boron concentration of the isolated loop to within the limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or c.

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 />.

l,

~

SURVEILLANCE REQUIREMENTS 1

4.4.1.5.1 At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, if required, verify that power ir/

removed from the valve operators.

4.4.1.5.2 At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, if required, verify that the boron i

concentration of an isolated loop is greater than or equal to the boron concentration of the operating loops.

i l

A loop is considered to be isolated when the hot and cold leg stop i

valves are both closed.

Three-loop operation is not allowed for Cycle 17.

j HADDAM NECK 3/4 4-9 Amendment No. J U, J M,

0117 l

q i

REACTOR COOLANT SYSTEM IDLED LOOP LIMITING CONDITION FOR OPERATION 3.4.1.8 The cold leg loop stop valve of an idled loop

• shall be shut and either: #

a.

The power removed from the valve operator, or b.

The boron concentration of the idled loop shall be maintained greater than or equal to the boron concentration of the operating loops.

APPLICABillTY: MODES I and 2.

l I

ACTION:

With the requirements of the above specification not satisfied, either:

l a.

Remove power from the valve operator within one hour, b.

Increase the boron concentration of the idled loop to within the limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or c.

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 />.

SURVEILLANCE RE0VIREMENTS 4.4.1.8.1 At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, if required, verify that power is removed from the valve operator.

4.4.1.8.2 At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, if required, verify that the boron concentration of an idled loop is greater than or equal to the boron concentration of the operating loops.

l

)

A loop is considered to be idled when the hot leg stop valve is open

)

and the cold leg stop valve is closed.

Three-loop operation is not allowed for Cycle 17.

HADDAM NECK 3/4 4-12 Amendment No. JEE, JfE, 0119

REACTOR COOLANT SYSTEM t

IDLED LOOP

]

LIMITING CONDITION FOR OPERATION i

3.4.1.9 The cold leg stop valve of an idled loop

• shall be shut and either:

a.

The power removed from the valve operator, or b.

The boron concentration of the idled loop shall be maintained i

greater than or equal to the boron concentration required to meet the SHUTDOWN MARGIN requirements of Specification 3.1.1.2 or 3.1.1.3 i

or the refueling boron concentration of Specification 3.9.1.

APPLICABILITY:

MODES 3, 4, 5, and 6 l

ACTION:

i With the requirements of the above specification not satisfied, either:

a.

Remove power from the valve operator within one hour, b.

Increase the boron concentration of the idled loop to within the limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or l

c.

Be in at least HOT SHUTDOWN 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 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SURVEILLANCE RE0UIREMENTS i

4.4.1.9.1 At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, if required, verify that power is removed from the valve operators.

4.4.1.9.2 At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, if required, verify that the boron concentration of an idled loop is greater than or equal to the boron concentration required to meet the SHUIDOWN MARGIN requirements of Specification 3.1.1.2 or 3.1.1.3 or the refueling boron concentration of Specification 3.9.1.

i A loop is considered to be idled when the hot leg stop valve is open and the cold leg stop valve is closed.

l HADDAM NECK 3/4 4-13 Amendment No. J/E, 0119

1 i

l REACTOR COOLANT SYSTEM

{

IDLED LOOP STARTUP l

LIMITING CONDITION FOR OPERATION i

3.4.1.11 A reactor coolant loop shall remain idled until:

a.

The temperature at the cold leg of the idled loop is within 20*F of the highest cold leg temperature of the operating loop (s),*

b.

The boron concentration of the idled loops is greater than or i

equal to the boron concentration required to meet the SHUTDOWN MARG:N requirements of Specification 3.1.1.2 or 3.1.1.3 or the refueling boron concentration of Specification 3.9.1.

APPLICABILITY:

MODES 3, o, 5, and 6 l

i ACTION:

l With the requirements of the above specification not satisfied, do not open

(

the idled loop cold leg stop valve.

i SURVEILLANCE j

i 4.4.1.11.1 The idled loop cold leg temperature shall be determined to be within 20*F of the highest cold leg temperature of the operating loop (s) within 30 minutes prior to opening the idled loop cold leg stop valve.

4.4.1.11.2 Within 30 minutes prior to opening the idled loop cold leg stop f

valve, the idled loop shall be determined to have a boron concentration greater than or equal to the bcron concentration required to meet the SHUTDOWN MARGIN requirements of Specification 3.1.1.2 or 3.1.1.3 or the j

refueling boron concentration of Specification 3.9.1.

If an idled loop is being started within 30 minutes after a reactor trip, this surveillance-requirement may be waived if the cold leg loop stop valve is closed for less than 15 minutes.

l 4.4.1.11.3 At least once per refueling outage the stop valve / temperature i

t interlock shall be determined operable by verifying that the cold leg stop valve does not open if the cold leg temperature in the loop is more than 20'F cooler than the highest temperature of the remaining operating loops.

j 4.4.1.11.4 At least once per refueling. outage the reactor coolant pump, loop stop and bypass valve interlock operability shall be demonstrated.

l An operating loop (s) may be a Reactor Coolant loop (s) or a Residual Heat Removal loop (s).

HADDAM NECK 3/4 4-15 Amendment No. J D,

0120 j

3)

Prior to returning the valve to service following maintenance, repair, or replacement work on the valve, and 4)

Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following valve actuation due to flow through the valve.

The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4 for Specification h, above.

i.

In addition to surveillance requirement 4.4.6.2.1.g, at least once per refueling outage, perform an operational leak rate test for those portions of the HPSI, charging and RHR systems outside of containment used for or pressurized during recirculation (with the exception of RHR suction piping). The test shall be conducted at a hydrostatic pressure corresponding to the operating pressure under accident conditions. The following provides the alternate testing for the RHR suction piping:

1.

Containment Sump to RH-MOV-22/RH-V-808A -

Test for leakage during the normally scheduled ILRT.

l 2.

RH-MOV-22 to RH-CV-783 and RH-V-808A to RH-CV-808A -

Piping to be tested at a pressure of approximately 6 psi. The leak rate will be extrapolated to the operating pressure under accident conditions.

3.

Piping Downstream of RH-CV-783 and RH-CV-808A -

Piping to be tested at approximately 30 psi. The leak rate will be extrapolated to the operating pressure under accident conditions.

• • Except for those portions of the HPSI, Charging and RHR suction piping which are not testable at accident pressure during normal operation, as defined below.

liPSI System - Those portions of HPSI suction piping downstream of the HPSI suction valves (SI-MOV-854A and B) and RHR/HPSI Crosstie valves (SI-MOV-901 and 902) and upstream of the HPSI pump suctions.

CHARGING SYSTEM - Those portions of charging suction piping downstream of the RHR/ Charging Crosstie Valves (RH-MOV-33A and B) and upstream of the charging pump suctions.

RHR SYSTEM - Those portions of the RHR suction piping between the containment sump and the RHR pump suctions.

The above piping will be tested in accordance with Specification 4.4.6.2.1.1 and also Specification 4.0.5.

HADDAM NECK 3/4 4-32a Amendment No. J/E, JJp, JEE, 0121

I SURVEILLANCE REQUIREMENTS (Continued)

L b.

If any periodic Type A test fails to meet 0.75 La, the test schedule for subsequent Type A tests shall_ be reviewed and approved by the Commission.

If two consecutive Type A tests fail to meet 0.75 La, a Type A test shall be performed at least every 18 months until two consecutive Type A tests meet 0.75 La at which r

time the above test schedule may be resumed or a corrective action plan may be prepared and submitted to the NRC that provides an acceptable alternative contingent on NRC approval.

c.

The accuracy of each Type A test shall be verified using the relationship:

(LTM + l - 0.25 L ) I b I IlTM + lo + 0.25 L,)

o a

c where:

L is the percent measured containment leakage per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at TM pressure P '

t L

is the percent superimposed leakage, g

L is the percent leakage obtained from the supplemental c

test result, and L,

is replaced with L f r reduced pressure tests.

t d.

Type B and C tests shall be conducted at intervals no greater than 24 months and at a pressure not less than Pa, 39.6 psig, using i

halogen gas detection, soap bubble, pressure decay, or other methods of equivalent sensitivity, except for tests involving:

1)

Air locks, and 2)

Purge supply and exhaust isolation valves with resilient material seals.

j e.

Air locks shall be tested and demonstrated OPERABLE by the i

requirements of Specification 4.6.1.3*

f.

Purge supply and exhaust isolation valves with resilient material seals shall be tested and' demonstrated OPERABLE by the requirements of Specification 4.9.9; g.

The provisions of Specification 4.0.2 are not applicable for Specifications 4.6.1.2.a through 4.6.1.2.d.

1 HADDAM NECK 3/4 6-3 Amendment No. J D, Jf),

oirz_

i ELECTRICAL POWER SYSTEMS i

SHUTDOWN f

LIMITING CONDITION FOR OPERATION 3.8.1.2 As a minimum, the following A. C. electrical power sources shall be OPERABLE:

a.

One circuit between the offsite transmission network and the Onsite Class IE Distribution System, and b.

One diesel generator, associated with the OPERABLE Onsite Class IE Distribution circuit, with:

1)

An engine-mounted fuel oil day tank containing a minimum volume of 400 gallons of fuel (except during engine operation),

2)

An underground fuel oil storage tank containing a minimum volume of 3,250 gallons of fuel,and 3)

A fuel transfer pump.

APPLICABILITY: MODES 5 and 6.

1 ACTION:

a.

With less than the above minimum required A.C. electrical power sources I

OPERABLE, immediately suspend all operations involving CORE ALTERATIONS, positive rer:tivity changes, movement of irradiated fuel, or crane operation with loads over the fuel storage pool. In addition, when in i

MODE 5 with less than two (2) steam generators OPERABLE, or in MODE 6 with the water level less than 23 feet above the reactor vessel flange, immediately initiate corrective action to restore the required sources to OPERABLE status as soon as possible.

1 b.

Entry into Mode 5 pursuant to Specification 3.0.4 with less than the minimum required A.C. electrical power sources OPERABLE is not permitted.

i SURVEILLANCE RE0VIREMENTS 4.8.1.2 The above required A.C. electrical power sources shall be demonstrated OPERABLE by the performance of each of the requirements of i

Specifications 4.8.1.1.1, 4.8.1.1.2 (except for Specification 4.8.1.1.2a.5, 4.8.1.1.2b, 4.B.1.1.2f), and 4.8.1.1.3.

i

]

f[fAMNECK 3/4 8-7 Amendment No. J/E, Jfy,

TABLE 4.8-2 l

BATTERY SURVEILLANCE REOUIREMENTS Weekly (I)

Quarterly (2)

I3)

PARAMETER LIMITS FOR EACH LIMITS FOR EACH ALLOWABLE DESIGNATED PILOT CONNECTED CELL VALUE FOR EACH CELL CONNECTED CELL Electrolyte

> Minimum level

>M... mum level Above 1:;, of Level indication mark, indication mark, pl ates, and < \\" above and < \\" above and not maximum level maximum level overflowing indication mark indication mark Float Voltage 22.10 volts 22.10 volts 22.07 volts Specifiy4, 21.200(5) 21.190 Not more than Gravity 0.020 below the 1

average of all connected cells average of all Average of all connected cells connec{gfcells

>1.200

>1.195 TABLE NOTATIONS (1)

For any Weekly parameter (s) outside the limit (s) shown, the battery may be considered OPERABLE provided that within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> all the Quarterly measurements are taken and found to be within their allowable values, and provided all Weekly and Quarterly parameter (s) are restored to within limits within the next 6 days.

(2) For any Quarterly parameter (s) outside the limit (s) shown, the battery may be considered OPERABLE provided that the Quarterly parameters are within their allowable values and provided the Quarterly parameter (s) are restored to within limits within 7 days.

(3) Any Quarterly parameter not within its allowable value indicates an inoperable battery.

(4) Corrected for electrolyte temperature and level.

(5) Or battery charging current is less than 2 amps when on charge.

HADDAM NECK 3/4 8-10 Amendment No. JM,

0124

REFUELING OPERATIONS I

3/4.9.2 INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.9.2.a As a minimum, two Source Range Neutron Flux Monitors shall be OPERABLE and operating, each with continuous visual indication in the control room and one with audible indication in the containment and control room when CORE ALTERATIONS or positive reactivity changes are taking place.

When CORE ALTERATIONS or positive reactivity changes are not taking place, at least one Source Range Neutron Flux Monitor shall be OPERABLE and operating with a visual indication in the control room and audible indication in the containment.

3.9.2.b As a minimum, two Source Range High Neutron Level Alarms i

(Containment Evacuation) shall be OPERABLE and operating with a minimum logic to audibly alarm in both the control room and containment of one (1) of two (2).

APPLICABILITY: MODE 6.

ACTION:

a.

With one of the above required monitors inoperable or not operating, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

b.

With both of the required monitors inoperable or not operating, determine the boron concentration of the Reactor Coolant System at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

SURVEILLANCE RE0VIREMENTS 4.9.2 Each Source Range Neutron Flux Monitor shall be demonstrated OPERABLE by performance of:

a.

A CHANNEL CHECK at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, b.

An ANALOG CHANNEL OPERATIONAL TEST within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior to the initial start of CORE ALTERATIONS, and c.

An ANALOG CHANNEL OPERATIONAL TEST at least once per 7 days.

i i

HADDAM NECK 3/4 9-2 Amendment No. JEE, Jf7 0125

REFUELING OPERATIONS 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE BUILDING l

LIMITING CONDITION FOR OPERATION 3.9.7 Loads in excess of 1650 pounds shall be prohibited from travel over fuel assemblies in the storage pool.

i APPLICABILITY: With fuel assemblies in the storage pool.

ACTION:

a.

With the requirements of the above specification not satisfied, place the crane load in a safe condition.

t b.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REOUIREMENTS 4.9.7 Administrative controls that prevent the travel of loads in excess of 1650 pounds over fuel assemblies shall be in place prior to lifting a load in excess of 1650 pounds.

i t

'I HADDAM NECK 3/4 9-7 Amendment No. J U,

0126

1/4.1 REACTIVITY CONTROL SYSTEMS BASES-3/4.1.1 BORAT10N CONTROL 3/4.1.1.1, 3/4.1.1.2. 3/4.1.1.3, and 3/4.1.1.4 SHUTDOWN MARGIN l

A sufficient SHUTDOWN MARGIN ensures that: (1) the reactor can be made subcritical from all operating conditions, (2) the reactivity transients associated with postulated accioent conditions are controllable within acceptable limits, and (3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel The most restrictive depletion, RCS boron concentration, and RCS T @o.

condition during MODES 1, 2 and 3 occurs at ell f-cycle life (E0L), and is associated with a postulated steam line break accident and resulting RCS cooldown.

In the accident analysis, a minimum SHUTDOWN MARGIN of 1800 pcm for four loop operation and 2600 pcm for three loop operation is assumed.

Operation in MODE 3 with two operating reactor coolant pumps is bounded by the four loop steam line break analysis. Operation in MODE 3 with one operating reactor coolant pump and two OPERABLE reactor coolant loops (both loop stop valves open in each loop) is bounded by the three loop steam line break analysis.

Because of the short time involved, the 2600-pcm SHUTDOWN MARGIN limit need not be applied to the closure of the cold leg stop valve in order to restart the reactor coolant pumps from an initial four loop operation condition.

The most restrictive condition in MODES 4 and 5 is associated with the boron dilution accident.

In the analysis of this accident, a minimum SHUTDOWN MARGIN of 3100 pcm in MODES 4 and 5 is required to control the reactivity transient, Accordingly, the SHUTDOWN MARGIN requirements are based upon this limiting condition and are consistent with j

current design basis assumptions.

3/4.1.1.5 MODERATOR TEMPERATURE COEFFICIENT l

The limits on the moderator temperature coefficient (MTC) are provided to ensure that the value of this coefficient remains within the limiting condition assumed in the accident and transient analysis.

The MTC values of this specification are associated with a specific set of plant conditions; measurement of MTC values at conditions other than those explicitly stated with extrapolation to the specified conditions is 3

acceptable. Correction factors shall account for fuel and moderator temperature and boron concentration.

f HADDAM NECK B3/4 1-1 Amendment No. JJE, JfE, 0127

O REACTIVITY CONTROL SYSTEMS j

BASES MODERATOR TEMPERATURE COEFFICIENT (Continued) i The Surveillance Requirement for measurement of the MTC at the beginning of the fuel cycle is adequate to confirm that the MTC remains within its limits since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup.

3/4.1.1.6 MINIMUM TEMPERATURE FOR rRITICALITY l

j This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 525'F. This i

limitation is required to ensure: (1) the moderator temperature coefficient 1s within it analyzed temperature range, (2) the trip instrumentation is within its normal operating range, (3) the pressurizer is capable of being in an OPERABLE status with a steam bubble, and (4) the reactor vessel is above its minimum RT temperature.

NDT 3/4.1.2 B0 RATION SYSTEMS The boration systems ensure that negative reactivity control is available during each MODE of facility operation. The components required to perform this function include: (1) borated water sources, (2) charging pumps, (3) separate flow paths, (4) boric acid transfer pumps, (5) associated Heat Tracing Systems, and (6) an emergency power sun 91y from OPERABLE diesel generators.

With the RCS average temperature above 200*F a minimum of two boron injection flow paths are required to ensure single functional capability in the event an assumed failure renders one of the flow paths inoperable..The i

boration capability of either flow path is sufficient to provide the required SHUTDOWN MARGIN of 3100 pcm from expected operating conditions after xenon decay and cooldown to 200*F. The maximum expected boration capability requirement occurs at E0L from full power equilibrium xenon conditions, and the minimum required volume of 12,000 gallons of 14,000-ppm borated water from the boric acid tank meets this requirement.

With the RCS temperature below 200*F, one boration 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 boration system becomes inoperable.

HADDAM NECK B3/4 1-2 Amendment No. JJE, JfE, C127 1

)

.j

t 3/4.2 POWER DISTRIBUTION LIMITS

. BASES i

i The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequen-cy) events by:

(I) maintaining the minimum DNBR in the core greater than or equal to I.30 during normal operation and in short-term transients, and (2) limiting the fission gas release, fuel pellet temperature, and cladding 1

mechanical properties to within assumed design criteria.

In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS Interim Acceptance Criterion limit of 2300*F peak cladding temperature for stainless steel clad fuel and the 10CFR50.46 and Appendix K limit of 2200*F peak cladding temperature for zircaloy fuel are not exceeded.

3/4.2.1 AXIAL OFFSET The AXIAL OFFSET specification provides continuous confirmation of accept-able LINEAR HEAT GENERATION RATES (LHGR) during the time interval between incore measurements.

3/4.2.2 LINEAR HEAT GENERATION RATE Limiting the peak LINEAR HEAT GENERATION RATE (LHGR) during Condition I events provides assurance that-the initial condition assumed for LOCA analyses are met and the peak cladding temperature limits are not exceeded.

N 3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR F,,

w ThelimitontheNUCLEARENTHALPYRISEHOTCHANNELFACTOR(FfH)ens the minimum DNBR limit is not exceeded.

N The F is measurable, but will normally only be determined periodically as

'j specihYedinSpecification4.2.3.I.2and4.2.3.2.2.

This periodic surveillance is sufficient to insure that the limits are maintained provided:

a.

The control rod insertion limits provided in the TECHNICAL REPORT SUPPORTING CYCLE OPERATION are maintained, and b.

The AXIAL OFFSET limits provided in the TECHNICAL REPORT SUPPORTING

]

CYCLE OPERATION are maintained.

)

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

The full power limits include a 4% incore measurement uncertainty.

3/4.2.4 00ADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial _ power distribu-tion satisfies the design values used in power capability analysis.

Radial HADDAM NECK B3/4 2-1 Amendment No. J2E, JfE, 0120

~

=

j INSTRUMENTATION BASFS 3.4.3.3.6 FIRE DETECTION INSTRUMENTATION (Continued) l equipment and is an integral element in the overall facility Fire Protection Program.

Fire detectors that are used to actuate Fire Suppression Systems represent a more critically important component of a plant's Fire Protection Program than detectors that are installed solely for early fire warning and notifi-I cation.

Consequently, the minimum number of OPERABLE fire detectors must be greater.

The loss of detection capability for Fire Suppression Systems, actuated by fire detectors, represents a significant degradation of fire protection for any area. As a result, the establishment of a fire watch patrol must be initiated at an earlier stage than would be warranted for the loss of -

detectors that provide only early fire warning. The establishment of fre-quent fire patrols in the affected areas is required to provide detection capability until the inoperable instrumentation is restored to OPERABILITY.

3/4.3.3.7 RADI0 ACTIVE LIOUID EFFLUENT MONITORING INSTRUMENTATION The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The Alarm / Trip.Setpoints for these instruments shall be calculated and adjust 2d in accordance with the methodology and parameters in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20.

7 The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

3/4.3.3.8 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION The radioactive gaseous effluent instrumentation is provided to monitor and f

control, as applicable, the releases of radioactive materials in gaseous i

effluents during actual or potential releases of gaseous effluents. The Alarm / Trip Setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the REM 0DCM to ensure-that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is consis-tent with the requirements of General Design Criteria 60, 63, and 64 of j

Appendix A to 10 CFR Part 50.

3/4.3.3.9 BORON DILUTION ALARM j

The shutdown monitors provide indication of positive reactivity insertion during operation in Modes 3, 4, 5, and 6.

The indication is credited in the l

Boron Dilution design basis analysis.

l 1

l HADDAM NECK B3/4 3-3 Amendment No. J U,

ons

t INSTRUMENTATION i

BASES 3/4.3.4 INTERNAL FLOOD PROTECTION l

The liquid level instrumentation is provided to monitor _ liquid levels in the i

areas of potential flooding caused by local pipe ruptures. The system ensures t

that early warning will occur so that protective action can be taken in the event of a localized flooding condition in areas of the plant that house safety-related equipment. The loss of detection capability represents a degradation of flooding protection for any area. As a result, the establishment of a liquid level watch patrol must be initiated at an early stage. The establishment of frequent liquid level watch patrols in the affected areas is required to provide detection capability until the inoper-i able instrumentation is restored to OPERABILITY.

i 1

i i

i l

t i

HADDAM NECK B3/4 3-4 Amendment No. J7E, 0129 i

l REACTOR COOLANT SYSTEM BASES l

3/4.4.1 REACTOR COOLANT SYSTEM LOOPS AND COOLANT CIRCULATION (Continued) l l

i The restrictions on starting an RCP with one or more RCS cold legs less than or equal to 315*F are provided to prevent RCS pressure transients, caused by

{

energy additions from the Secondary Coolant System, which could exceed the t

limits of Appendix G to 10 CFR Part 50. The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by' restricting starting of the RCPs to when the secondary water temperature of l

each steam generator is less than 20*F above each of the RCS cold leg 5

temperatures.

The requirement to maintain the boron concentration of an isolated / idled i

loop greater than or equal to the boron concentration of the operating loops or the boron concentration required to meet SHUTDOWN MARGIN requirements ensures that no unacceptable reactivity addition to the core could occur during startup of an isolated / idled loop. Verification of the baron concen-

[

tration in an isolated / idled loop prior to opening the stop valves provides a reassurance of the adequacy of the boron concentration in the isolated / idled loop.

Startup of an isolated / idled loop could inject cool water from the loop into

{

the core. The reactivity transient resulting from this cool water injection is minimized by prohibiting isolated / idled loop startup until its tempera-ture is within 20*F of the operating loops.

j t

3/4.4.2 SAFETY VALVES l

i The pressurizer Code safety valves operate to prevent the RCS from being l

pressurized above its Safety Limit of 2735 psig. The required relieving capacity of each safety valve is 240,000 lbs. per hour at 2,485 psig as i

assumed in the safety analysis. Each safety valve is conservatively i

designed to relieve 293,300 lbs. per hour of saturated steam at 2485 psig.

The relief capacity of a single safety valve is adequate to relieve ary overpressure condition which could occur during shutdown.

In the event that no safety valves are OPERABLE, an operating RHR loop, connected to the RCS, provides overpressure relief capability and will prevent RCS overpressur-l ization.

In addition, the Overpressure Protection System provides a diverse i

I means of protection against RCS overpressurization at low temperatures.

During operation, all pressurizer Code safety valves must be OPERABLE to prevent the RCS from being pressurized above its Safety Limit of 2735 psig.

The combined relief capacity of all of these valves is greater than the j

maximum surge rate resulting from a complete loss-of-load assuming no i

Reactor trip until the first Reactor Trip System Trip Setpoint is reached (i.e., no credit is taken for a direct Reactor trip on the loss-of-load) and also assuming no operation of the power-operated relief valves or steam dump l

valves.

Demonstration of the safety valves' lift settings will. occur only during i

shutdown and will be performed in accordance with the provisions of j

Section XI of the ASME Boiler and Pressure Code.

't I

f HADDAM NECK B3/4 4-2 Amendment No. R5, 0131

?

j.

CONTAINMENT SYSTEMS BASES 3/4.6.1.5 AIR TEMPERATURE (Continued) and a main steam line break inside the containment. Measurements shall be taken from all OPERABLE temperature detectors to determine the average air temperature.

I 3/4.6.1.6 CONTAINMENT VESSEL STRUCTURAL INTEGRITY l

k This limitation ensures that the structural integrity of the containment will be maintained comparable to the original design standards for the life of the facility. Structural integrity is required to ensure that the i

containment will withstand the maximum pressure of 39.6 psig in the event of a LOCA. A visual inspection in conjunction with the Type A leakage tests is sufficient to demonstrate this capabil!'.y.

3/4.6.1.7 CONTAINMENT VENTILATION SYSTEM l

The 42-inch containment purge supply and exhaust isolation valves and the l

8-inch bypass valve are required to be closed and locked closed during plant l

operation since these valves have not been demonstrated capable of closing l

during a LOCA or steam line break accident. Maintaining these valves locked l

closed during plant operations ensures that excessive quantities of radio-l active materials will not be released via the Containment Purge System. To I

l provide assurance that these containment valves cannot be inadvertently l

opened, the valves are locked closed in accordance with Standard Review Plan l

6.2.4 which includes mechanical devices to lock the valve closed.

Containment post accident hydrogen venting can be accomplished by two methods. One uses the containment air particulate monitoring system and the other uses the containment purge exhaust system. These methods are not required in any short time frame after an accident; it is expected that months may elapse.

In any event, if the systems are not operable because of maintenance reasons, they can be made operable. System operability can be readily obtained provided access into the containment is not required.

Containment purge is utilized as a back-up means of venting hydrogen from j

the containment following a loss-of-coolant accident. The containment air particulate monitoring system provides the primary means of purging because it provides adequate purge flow to prevent an explosive mixture build-up while allowing fine control of the release of radioactivity during purges.

When necessary to effect repairs to the containment purge or purge bypass isolation valves, a blank flange must be applied to the 42" purge air exhaust penetration inside the reactor containment so that the containment remains leak tight. This renders the purge system inoperable for a finite 1

time. Seven days is considered a reasonable length of time for repair parts to be received, installed and the system retested for leak tightness and returned to service.

l t

HADDAM HECK B3/4 6-?

Amendment No. J U,

l 0130 v

l 3/4.7 PLANT SYSTEMS BASES l

3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line Code safety valves ensures that the Secondary Coolant System pressure will be limited to below 110%, (1100 l

psia), of its design pressure of 1000 psia during the most severe anticipated system operational transient. The maximum relieving capacity is j

l associated with a Turbine trip from 100% RATED THERMAL POWER coincident with l

an assumed loss of condenser heat sink (i.e., no steam bypass to the condenser).

The specified valve lift settings and relieving capacities are in accordance with the requirements of Section XI of the ASME Boiler and Pressure Vessel Code, 1971 Edition. The design total relieving capacity for all valves on all of the steam lines is 9,504,000 lbs/hr which is 120% of the total secondary steam flow of 7,872,000 lbs/hr at 100% RATED THERMAL POWER.

3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM The OPERABILITY of the Auxiliary Feedwater System ensures that the Reactor Coolant System can be cooled down to less than 350*F from normal operating conditions in the event of a total loss of offsite power.

Each steam turbine-driven auxiliary feedwater pump has a capacity sufficient l

to ensure adequate delivery of feedwater flow to remove decay heat and l

reduce the Reactor Coolant System temperature to less than 350*F within the Residual Heat Removal System operating range. With one auxiliary feedwater i

pump inoperable, the safest mode of operation is HOT SHUTDOWN with the decay heat removal function capable of being provided by the RHR System. With two steam turbine-driven feedwater pumps inoperable, at least one pump must be restored to OPERABLE with 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from the time that the second pump is declared inoperable, or be in HOT STANDBY within the next six hours and in HOT SHUTDOWN with the following six hours.

In addition, both the pumps must be restored to OPERABLE within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> from the time of initial loss of the first pump or be in HOT STANDBY in the next six haurs and HOT SHUTDOWN

[

within the following six hours.

The auxiliary feedwater (AFW) system's design basis requires AFW to be automatically initiated and to be independent of any AC electrical power scurce for at least two hours. The AFW pump / turbine governor's DC powered hydraulic pump, controls, and DC power supply are required to be OPERABLE for the associated AFW pump to be OPERABLE.

If the DC pump automatic start instrumentation does not function, the associated AFW pump remains OPERABLE as long as the DC powered hydraulic pump is started and maintained operating in accordance with the stated ACTION statement.

3/4.7.1.3 AUXILIARY FEEDWATER SUPPLY The OPERABILITY of the demineralized water storage tank (DWST) and primary water storage tank (PWST) with the minimum water volume ensures that l

sufficient water is available to maintain the RCS at HOT STANDBY conditions l

for 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> with steam discharge to the atmosphere concurrent with total r

HADDAM NECK B3/4 7-1 Amendment No. JJE, JfE, JfE, 0133

PLANT SYSTEMS BASES AUXILIARY FEEDWATER SUPPLY (Continued) loss-of-offsite power. The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.

In addition, the auxiliary feedwater system can be in5tiated manually.

In this case, feedwater is available from the DWST by gravity feed to the auxiliary feedwater pump. The specified 50,000 gallons of water in the DWST is adequate for decay heat removal for a period of at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Within this period, decay heat removal demands are reduced to approximately 150 gpm. Makeup water is available during this period from the PWST which contains a minimum volume of 80,000 gallons. The PWST transfer pumps can transfer 200 gpm from the PWST to the DWST. An alternate supply can be provided from the 100,000 gallons Recycled Primary Water Storage Tank.

3/4.7.1.4 SPECIFIC ACTIVITY The limitations on Secondary Coolant System specific activity ensure that the resultant offsite radiation dose will be limited to a small fraction of 10 CFR Part 100 dose guideline values in the event of a steam line rupture.

This dose also includes the effects of a coincident 0.4 gpm reactor-to-secondary tube leak in the steam generator of the affected steam line.

These values are consistent with the assumptions used in the safety analyses.

3/4.7.1.5 MAIN STEAM LINE TRIP VALVES

}

The OPERABILITY of the main steam line trip valves ensures that no more than one steam generator will blowdown in the event of a steam line rupture.

t This restriction is required to: (1) minimize the positive reactivity effects of the Reactor Coolant System cooldown associated with the blowdown, and (2) limit the pressure rise within containment in the event the steam line rupture occurs within containment. The OPERABILITY of the main steam t

line trip valves within the closure times of the Surveillance Requirements are consistent with the assumptions used in the safety analyses.

l 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION The limitation on steam generator pressure and temperature ensures that the pressure-inducedstressesinthesteamgeneratorsdonotexceedthgmaximum allowable fracture toughness stress limits. Thglimitationsof70Fand200 psig are based on a steam generator RTNDT of 10 F and are sufffcient to prevent brittle fracture. The heatup and cooldown rate of 100 F/hr for the steam generators are specified to ensure that stresses in these vessels are maintained within acceptable design limits.

3/4.7.3 SERVICE WATER SYSTEM The OPERABILITY of the Service Water System ensures that sufficient cooling capacity is available for continued operation of safety-related equipment during normal and accident conditions. The two service water pumps which l

HADDAM NECK B3/4 7-2 Amendment No. JfE, J,45 0133

q PLANT SYSTEMS i

BASES i

3/4.7.3 SERVICE WATER SYSTEM (Continued) e are powered by the "A" EDG must be operable during the construction period.

The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the safety analysis. A service water header is comprised of the two service water pumps associated with each diesel generator and the I

L l

l i

t i

L liADDAM NECK B3/4 7-2a Amendment No.

0133 i

f f

O I

l l,

l PLANT SYSTEMS I

l BASES 3/4.7.9 FEEDWATER ISOLATION VALVES The accident analysis for a main steam line break assumes that the main feedwater isolation valves will close on a containment isolation actuation signal (CIAS). Also, the closure of these valves based on a CIAS is credited in determining the Pressure / Temperature limits for the purpose of environmental qualification. The feedwater isolation valves act as a backup to the feedwater regulation valves in the event a feedwater regulation valve fails open during a Main Steam Line Break.

3/4.7.10 EXTERNAL FLOOD PROTECTION The thresholds regarding flood protection ensure that facility protective actions will be taken (and the orderly shutdown of the plant to MODE 3 will be made) in the event of flood conditions. The estimated Connecticut River probable maximum flood (PMF) level, including wave effects (i.e., still water level), is 39.5 feet mean sea level. Normal flood control measures provide 1'

protection to safety-related equipment to El. 30 feet mean sea level. Normal flood protection to this elevation is based on a low probability of exceedance and-structural capacity limitations. Based on the one to two day rise period of the PMF, alternative means of providing decay heat removal for flooding events up to the PMF is provided in A0P 3.2-24.

3/4.7.11 PRIMARY AUXILIARY BUILDING AIR CLEANUP SYSTEM l

1 PAB Air Cleanup System consists of two exhaust fans, two prefilters, a HEPA-HECA filter assembly, and interconnecting ductwork.

Air cleanup is accomplished using one exhaust fan, one prefilter, the HEPA-HECA filter, and interconnected ductwork.

The radiological consequences analyses for loss-of-coolant accidents assume Primary Auxiliary Building efficiencies which are ensured by this Technical Specification. Also, in consideration of a fuel handling accident inside containment, (i.e., when the containment is being purged) the purge discharge would be directed through the Primary Auxiliary Building charcoal filters.

Credit is again taken for these filters in reducing the radiological consequences.

i HADDAM NECK B3/4 7-6 Amendment No. J/E, 0134

4 3/4.8 ELECTRICAL POWER SYSTEMS f

BASES 3/4.8.1. 3/4.8.2 AND 3/4.8.3 A. C. SOURCES D. C. SOURCES. ONSITE POWER f

DISTRIBUTION The OPERABILITY of the A. C. and D. C. power sources and associated distribution systems during operation ensures that sufficient power will be available to supply the safety-related equipment required for: (1) the safe shutdown of the facility, and (2) the mitigation and control of accident conditions within the facility. The minimum specified independent and redundant A.C. and D.C. power sources and distribution systems satisfy the requirements of General Design Criterion 17 of Appendix A to 10 CFR Part 50.

The ACTION requirements specified for the levels of degradation of the power sources provide restriction upon continued facility operation commensurate i

with the level of degradation. The OPERABILITY of the power sources is consistent with the initial condition assumptions of the safety analyses and based upon maintaining at least one redundant set of onsite A.C. and D.C.

L power sources and associated distribution systems OPERABLE during accident conditions coincident with an assumed loss-of-offsite power and single i

failure of the other onsite A.C. source. When one diesel generator is inoperable, there is an additional ACTION requirement to verify that the charging pump, HPSI pump, LPSI pump and RHR pump that depend on the remaining OPERABLE diesel generator as a source of emergency power, are also

?

OPERABLE. This requirement is intended to provide assurance that a loss-of-offsite power event will not result in a complete loss of safety function of critical systems during the period one of the diesel generators is inoperable. The term, verify, as used in this context means to administratively check by examining logs or other information to determine if certain components are out-of-service for maintenance or other reasons.

It does not mean to perform the Surveillance Requirements needed to r

demonstrate the OPERABILITY of the component.

The OPERABILITY of the minimum specified A.C. and D.C. power sources and associated distribution systems during shutdown and refueling condition ensures that:

(1) the facility can be maintained in the refueling or shutdown condition for extended time periods, and (.2) sufficient instrumentation and control capability is available for monitoring and maintaining'the facility status.

The Surveillance Requirements for demonstrating the OPERABILITY of the i

diesel generators are based on the recommendations of Regulatory Guides 1.9,

" Selection of Diesel Generator Set Capacity for Standby Power Supplies",

March 10,1971; 1.108, " Periodic Testing of Diesel Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants", Revision 1, August i

1977; and 1.137, " Fuel-Oil Systems for Standby Diesel Generators", Revision i

1, October 1979, and guidance given in Generic Letter 84-15.

HADDAM NECK B3/4 8-1 Amendment No. J2E, D135

l i

REFUELING OPERATIONS l

l BASES 3/4.9.6 MANIPULATOR CRANE j

t The OPERABILITY requirements for the manipulator cranes ensure that:

(1) manipulator cranes will be used for movement of control rod drive shafts and fuel assemblies, (2) each crane has sufficient load capacity to lift a drive shaft or fuel assembly, and (3) the core internals and reactor vessel are i

protected from excessive lifting forces in the event they are inadvertently engaged during lifting operations.

3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE BUILDING l

l The restriction on movement of loads in excess of the nominal weight of a fuel and control rod assembly and associated handling tool over other fuel i

assemblies in the storage pool ensures that in the event this load is dropped: (1) the activity release will be limited to that contained in 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 i

with the activity release assumed in the safety analysis.

3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION l

The requirement that at least one RHR LOOP be in operation ensures that: (1) sufficient cooling capacity is available go remove decay heat and maintain the water in the reactor vessel below 140 F as required during the REFUELING MODE, 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 l

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 23 feet i

of water above the reactor vessel flange, a large heat sink is available for i

core cooling. Thus, in the event of a failure of the operating RHR LOOP, t

adequate time is provided to initiate emergency procedures to cool the core.

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

The OPERABILITY of this system is required to restrict the j

release of radioactive material from the containment atmosphere to the environment.

i l

l HADDAM NECK B 3/4 9-2 Amendment No. JL5 l

0136

{

i I

.i 3/4.10 SPECIAL TEST EXCEPTIONS BASES 3/4.10.1 SHUTDOWN MARGIN 3

This Special Test Exception provides that a minimum amount of control rod worth is immediately available for reactivity control when tests are t

performed for control rod worth measurement. This special test exception is required to permit the periodic verification of the actual versus predicted j

core reactivity condition occurring as a result of fuel burnup or fuel cycling operations.

3/4.10.2 PHYSICS TESTS This Special Test Exception permits PHYSICS TESTS to be performed at less than or equal to 5% of RATED THERMAL POWER with the RCS T slightly lower than normally allowed so that the fundamental nuclear chaf5Eteristics of the i

core and related instrumentation can be verified.

In order for various characteristics to be accurately measured, it is at times necessary to 1

operate outside the normal restrictions of these Technical Specifications.

For instance, to measure the moderator temperature coefficient at BOL, it is necessary to position the various control rods at heights which may not normally be allowed by Specification 3.1.3.6 which in turn may cause the RCS T

to fall slightly below the minimum temperature of Specification 3.Y9.4.

l a

1 3/4.10.3 POSITION INDICATION SYSTEM-SHUTDOWN This Special Test Exception permits the Position Indication Systems to be inoperable during rod drop time measurements.

The exception is required since the data necessary to determine the rod drop time are derived from the l

induced voltags in the position indicator coils as the rod is dropped. This induced voltage is small compared to the normal voltage and, therefore, cannot be observed if the Position Indication Systems remain OPERABLE.

3/4.10.4 POSITION INDICATION SYSTEM - OPERATING l

This Special Test Exception permits the IRPI system to be inoperable during the performance of data collection / verification / adjustment testing of the IRPI. The testing is required to develop and implement correction factors for each individual rod position indicator. While the IRPI system is inoperable, the indicated individual rod position cannot be used to verify control rod alignment (Specification 3.1.3.1) or control rod insertion limits (Specifications 3.1.3.5, 3.1.3.6.1 and 3.1.3.6.2).

The actual rod i

position for banks, C, D and A is, however, unaffected by this testing.

i i

q HADDAM HECK B3/4 10-1 Amendment No. J/E, J/7, D137

+

o 3/4.11 RADI0 ACTIVE EFFLUENTS BASES 3/4.11.1 LIOUID EFFLUEFTS 3/4.11.1.1 CONCENTRATION This specification is provided to ensure that the concentration of radioactive materials released in liquid waste effluents from the site will be less than the concentration levels specified in 10 CFR Part 20, Appendix B, Table II, Column 2.

This limitation provides additional assurance that the levels of radioactive materials in bodies of water outside the site will result in exposures within:

(1) the Section II.A design objectives of Appendix I,10 CFR Part 50, to a MEMBER OF THE PUBLIC, and (2) the limits of 10 CFR 20.106(e) to the population. The concentration limit for dissolved or entrained noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water'using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.

3/4.11.1.2 DOSE. l.IOUIDS l

This specification is provided to implement the requirements of Sections II.A, III.A, and IV.A of Appendix 1, 10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in-Section II.A of Appendix 1.

The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid effluents will be kept "as low as is reasonably achievable". The dose calculation methodology and parameters in the REMODCM implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I is to be shown by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the REM 0DCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents 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 Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977, and Regulatory Guide 1.113, " Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," April 1977.

HADDAM NECK B3/4 11-1 Amendment No. J U,

0138

4 6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The Vice President - Paddam Neck shall be responsible for overall l

I facility operation and shall delegate, in writing, the succession to this responsibility during his absence.

l 6.2 ORGANIZATION 6.2.1 ONSITE AND OFFSITE ORGANIZATIONS 7

Onsite and offsite organizations shall be established for unit operation and' corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting the safety of the nuclear power plant.

a.

Lines of authority, responsibility, and communication shall be established and defined for the highest management levels through intermediate levels to and including all operating organization positions. These relationships shall be documented and updated, as appropriate, in the form of organization charts, functional i

descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements shall be documented in i

the Quality Assurance Topical Report.

b.

The Vice President - Haddam Neck shall be responsible for overall l

unit safe operation and shall have control over those onsite activities necessary for safe operation and maintenance of the pl ant.

c.

The Executive Vice President-Nuclear, shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety.

d.

The individuals who. train the operating staff and those who carry out health physics and quality assurance functions may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their independence from operating pressures.

6.2.2 FACILITY STAFF a.

Each on-duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.2-1; b.

At least one licensed Operator shall be in the control room when fuel is in the reactor.

In addition, while the facility is in MODE 1, 2, 3 or 4, at least one licensed Senior Operr. tor shall be in the control room; HADDAM NECK 6-1 Amendment No. J/E, JEE, 0139

O i

7 s.

ADMINISTRATIVE CONTROLS f.

Successful completion of the Thames Valley State Technical College associate's degree in Nuclear Engineering Technology program, provided that the individual was enrolled in the program by October 1, 1987.

2.

Dedicated STA:

Must meet the STA training criteria of NUREG-0737, Item I.A.1.1, and have received specific training in plant design, and response and analysis of the plant for transients and j

accidents.

6.4 TRAINING 6.4.1 A retraining and replacement training program for the facility staff shall be maintained under the direction of the Nuclear Unit Director and shall meet or exceed the requirements and recommendations of Section 5.5'of ANSI N18.1-1971 and 10CFR55.59. The Director-Nuclear Training has the overall respor.sibility for the implementation of the Training Program.

I 6.4.2 A training program for the Fire Brigade shall be maintained under the direction of the Director-Nuclear Training and shall meet or exceed the i

intent of Section 27 of the NFPA Code-1975, except for Fire Brigade training sessions which shall be held at least quarterly.

i 6.5 REVIEW AND AUDIT 6.5.1 PLANT OPERATIONS REVIEW COMMITTEE (PORC)

FUNCTION t

l 6.5.1.1 The PORC shall function to advise the Vice President-Haddam Neck

[

on all matters related to nuclear safety.

4 COMPOSITION 6.5.1.2 The PORC shall be composed of the:

Chairperson:

Vice President - Haddam Neck l

Member:

Nuclear Unit Director J'

Member:

Operations Manager Member:

Maintenance Manager Member:

Instrument and Contro_1 Manager Member:

Reactor Engineer l

Member:

Engineering Manager Member:

Nuclear Services Director Member:

Plant Quality Services Supervisor Member:

Chemistry Manager I

Member:

Health Physics Manager l

Member:

Security Manager i

i HADDAM NECK 6-5 Amendment No. J S, JJJ, JEJ, j

014D

ADMINISTRATIVE CONTROLS s

-r AtTERNATES 1

1 6.5.1.3 All alternate members shall be appointed in writing by the PORC Chairperson to serve on a temporary basis; however, no more than two alternates shall participate as voting members in PORC activities at any one i

time.

MEETING FRE0VENCY 6.5.1.4 The PORC shall meet at least once per calendar month and as convened by the PORC Chairperson or his/her designated alternate.

OUORUM 6.5.1.5 The quorum of the PORC shall consist of the Chairperson or his/her designated alternate and four members including alternates.

RESPONSIBILITIES 6.5.1.6 The PORC shall be responsible for:

a.

Review of: (1) all procedures required by Specification 6.8 and changes thereto, and 2) any other proposed procedures or changes thereto as determined by the Vice President - Haddam Neck to l

l affect nuclear safety; 4

b.

Review of all proposed tests and experiments that affect nuclear safety; c.

Review of all proposed changes to the Technical Specifications; d.

Review of all proposed changes or modifications to plant systems or equipment that affect nuclear safety; e.

Investigation of all violat: ens of the Technical Specifications, including the preparation and forwarding of reports covering evaluation and recommendations to prevent recurrence, to the Executive Vice President-Nuclear and to the Chairperson of the Nuclear Review Board; f.

Review of all REPORTABLE EVENTS; g.

Review of facility operations to detect potential safety hazards; h.

Performance of special reviews, investigations or analyses and reports thereon as requested by the Chairperson of the Nuclear Review Board or the Vice President - Haddam Neck.

l 1

)

l 1

HADDAM NECK 6-6 Amendment No. JJE, JEE, 0140 d

e ADMINISTRATIVE CONTROLS i.

Review of the Security Plan and implementing procedures and shall i

submit recommended changes to the Chairperson of the Nuclear Review Board; j.

Review of the Emergency Plan and implementing procedures and shall submit recommended changes to the Chairperson of the Nuclear Review Board; AUTHORITY 6.5.1.7 The PORC shall:

a.

Report to and be advisory to the Vice President - Haddam Neck on l

those areas of responsibility specified in Section 6.5.1.6(a) through (j);

b.

Render determinations in writing to the Vice President - Haddam Neck if any item considered under Specification 6.5.1.6a. through d., above, as appropriate and as provided by 10CFR50.59 or 10CFR50.92 constitutes an unreviewed safety question or requires a significant hazards consideration determination.

c.

Provide written notification, meeting minutes may be used for this purpose, to the Executive Vice President-Nuclear and the Chairperson of the Nuclear Review Board of disagreement between the PORC and the Vice President - Haddam Neck; however, the Vice President - Haddam Neck shall have responsibility for resolution of such disagreements pursuant to Specification 6.1.1.

RECORDS 6.5.1.8 The PORC shall maintain written minutes of each meeting that, at a minimum, document the results of all PORC activities performed under the responsibility and authority provisions of these Technical Specifications.

A Copy shall be provided to the Chairperson of the Nuclear Review Board.

l 6.5.2 NUCLFAR REVIEW BOARD (NRB) t 00ALIFICATIONS 6.5.2.1 The minimum qualifications of NRB members are as follows:

a.

The Chairperson and NRB members shall have:

1.

an academic degree in engineering or physical science field, or hold a senior management position, and 2.

a minimum of five years technical experience in their d

respective field of expertise, and 3.

a minimum of nine (9) years combined academic and technical experience.

HADDAM NECK 6-7 Amendment No. J/J, JEE, 0140 7

t ADMlNISTRATTVE CONTROLS b.

The NRB shall collectively have the experience and competence required to review activities in the following areas:

1.

Nuclear power plant operations 2.

Nuclear engineering 3.

Chemistry and radiochemistry 4.

Metallurgy 5.

Nondestructive testing 6.

Instrumentation and control 7.

Radiological safety 8.

Mechanical and electrical engineering 9.

Administration 10.

Quality assurance practices COMPOSITTON 6.5.2.2 The NRB shall consist of no less than eight, nor more than eleven members including the Chairperson and the Vice President - Haddam Neck. The l

Chairperson and members of the NRB shall be appointed in writing by the Executive Vice President - Nuclear.

CONSULTANTS 6.5.2.3 Consultants shall be utilized as determined by the NRB Chairperson to provide expert advice to the NRB.

MEETING FRE0VENCY 6.5.2.4 The NRB shall meet at least once per 6 months.

OUORUM 6.5.2.5 The quorum of the NRB necessary for the nerformance of the NRB review and audit functions of these Technical Specifications shall consist of at least enough members to constitute a majority of the assigned members including the Chairperson or a designated alternate. No more than a minority of the quorum shall have line responsibility for operatio:t of the facility.

HADDAM NECK 6-8 Amendment No. JJE, JEE, 0140

e

=

ADMINISTRATIVE CONTROLS a.

The applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2, February 1978; b.

The requirements and recommendations of Sections 5.1 and 5.3 of ANSI N 18.7-1976.

c.

Fire Protection Program implementation.

d.

Quality controls for effluent monitoring, using the guidance ir, Regulatory Guide 1.21 Rev.1, June 1974.

e.

RADIOLOGICAL EFFLUENT MONITORING AND OFFSITE DOSE CALCULATION MANUAL (REMODCM) implementation except for Section I.E, Radiological Environmental Monitoring.

f.

PROCESS CONTROL PROGRAM implementation.

6.8.2 Each procedure of Specification 6.8.1, and changes thereto, shall be reviered by the PORC and shall be approved by the Vice President - Haddam Neck prior to implementation and reviewed periodically as set forth in each document or in administrative procedures.

6.8.3 Temporary changes to procedures of Specification 6.8.1 may be made provided:

a.

The intent of the original procedure is not altered; b.

The change is approved by two members of the plant management l

staff, at least one of whom holds a Senior Operator license on the unit affected; and c.

The change is documented, reviewed by the PORC and approved by the Vice President - Haddam Neck within 14 days of implementation.

l 6.8.4 Written procedures shall be established, implemented and maintained covering Section I.E., Radiological Environmental Monitoring, of the REMODCH.

6.8.5 All procedures and procedure changes required for the Radiological Environmental Monitoring Program of Specification 6.8.4 above shall be reviewed by an individual (other than the author) from the Radiological Assessment Branch or the Production Operation Services Laboratory (POSL) and e

approved by appropriate supervision.

l Temporary changes may be made provided the intent of the original procedure is not altered and the change is documented and reviewed by an individual (other than the author) from the Radiological Assessment Branch or the POSL, within 14 days of implementation.

.c HADDAM NECK 6-12 Amendment No. JJE, JEE, 0141 L

D ADMINISTRATIVE CONTROLS 6.13 RADIOLOGICAL EFFLUENT MONITORING AND OFFSITE DOSE CALCULATION MANUAL (REMODCH)

Section I,

Radiological Effluents Monitoring Manual, shall cutline the sampling and analysis programs to determine the concentration of radioactive materials released offsite as well as dose commitments to individuals in those exposure pathways and for those - radionuclides released as a result of facility operation.

It shall also specify operating guidelines for RADI0 ACTIVE WASTE TREATMENT SYSTEMS and report content.

Section II, the Offsite Dose Calculation Manual, shall describe the methodology and parameters to be used in the calculation of offsite doses-due to radioactive gaseous and liquid effluents and in the calculations of gaseous and liquid effluent monitoring instrumentation Alarm / Trip Setpoints consistent with the applicable LC0's contained in these Technical Specifications.

Changes to the REMODCH:

a.

Shall be documented and records of reviews performec shall be retained as required by Specification 6.10.3.m.

This documentation shall contain:

1)

Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change (s), and 2)

A determination that the change will maintain the level of radioactive effluent control required by 10 CFR 20.105, 40 CFR Part 190, 10 CFR 50.35a, and Appendix I to 10CFR Part 50 and not adversely impact the accuracy or reliability of effluent, dose, 'or setpoint calculations.

b.

Shall become effective after review and acceptance by PORC and the approval of the Vice President Haddam Neck.

l c.

Shall be submitted to the Commission in the form of a complete, legible copy of the entire REMM or ODCM, as appropriate, as a part of or concurrent with the Semiannual Radioactive Effluent Relecse Report for the period of the report in which any change was made.

Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g., month / year) the change was implemented.

6.14 RADI0 ACTIVE WASTE TREATMENT Procedures for liquid and gaseous radioactive effluent discharges from the facility shall be prepared, approved, maintained and adhered to for all operations involving offsite _ releases of radioactive effluents.

These procedures shall specify the use of appropriate RADI0 ACTIVE WASTE TREATMENT SYSTEMS utilizing the guidance provided in the REMODCM.

The Solid RADI0 ACTIVE WASTE TREATMENT SYSTEM shall be operated in accordance with the PROCESS CONTROL PROGRAM to process wet radioactive wastes to meet shipping and burial ground requirements.

HADDAM NECK 6-20 Amendment No. J/E, JEE, 0142

P c1r ADMINISTRATIVE CONTROLS 6.15 SYSTEMS INTEGRITY The licensee shall implement a program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. This program shall include the following:

a.

Provisions establishing preventive maintenance and periodic visual inspection requirements, and b.

Integrated leak test requirements for each system at a frequency not to exceed refueling cycle intervals.

6.16 PASS / SAMPLING AND ANALYSIS OF PLANT EFFLUENTS l

I The licensee shall implement and maintain a program which will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant gaseous effluents, and containment atmosphere samples under accident conditions. This program shall include the following:

a.

Training of personnel b.

Procedures for sampling and analysis, and c.

Provisions for maintenance of sampling and analysis equipment.

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HADDAM NECK 6-21 Amendment No. JJE, JEE, 0143