Text

Amend 158 to License DPR-61,making Editorial Changes to TSs Which Are Administrative in Nature
	ML20044H047
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	05/27/1993
From:	Stolz J; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20044H046	List: ML20044H045; ML20044H047;
References
	NUDOCS 9306070290
	Download: ML20044H047 (58)
	v • d • e

'#

o UNITED STATES i

g

,[

p, NUCLEAR REGULATORY COMMISSION i

n l

WASHINGTON, D. C. 20555 l

\\...../

\\

i l

CONNECTICUT YANKEE ATOMIC POWER COMPANY J

l DOCKET NO. 50-213 I

l HADDAM NECK PLANT AMENDMENT TO FACILITY OPERATING LICENSE bPR-61 e

o 1.

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

A.

The application for amendment by Connecticut Yankee Atomic Power i

Company (the licensee), dated March 16, 1993, complies with the standards and requirements of the_ Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set l

forth in 10 CFR Chapter I; B.

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

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

this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations-l D.

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

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

i l

9306070290 930527 PDR ADOCK 05000213 P

PDR

j

. l 2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to.this license amendment,

- i and paragraph 2.C.(2) of Facility Operating License No. DPR-61 is hereby amended to read as follows:

1 (2) Technical Snecifications The Technical Specifications contained in Appendix A, as revised through Amendment No.-I58, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of issuance, to be i

implemented within 30 days' of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION-j n. Stolz, Direct P oject Directorate 4

i ivision of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical l

Specifications Date of Issuance: May 27, 1993 j

r O O l

1 i

i

ATTACHMENT TO LICENSE AMENDMENT NO.158 FACILITY OPERATING LICENSE NO. DPR-61 DOCKET NO. 50-213

-i i

Replace the following pages of the Appendix A Technical Specifications with the enclosed pages.. The revised pages are identified by amendment number.and l

contain vertical lines indicating the areas of change.

Remove Insert VIII VIII-IX IX X

X XI XI l

XIII XIII XIV XIV XV XV XVIII XVIII XIX XIX l-2 1-2 3/4 2-7 3/4-2-7 3/4 3-2 3/4 3-2 3/4 3-3 3/4 3-3 3/4 3-5 3/4 3-5 3/4 3-7 3/4 3-7 3/4 3-12 3/4 3-12 s

3/4 3-17 3/4 3-17 3/4 3-18 3/4 3-18 3/4 3-24

3/4 3-24

+

3/4 3-43 3/4 3-43 3/4 4-9 3/4 4-9 3/4 4-12 3/4 4-12 3/4 4-13 3/4 4-13 d

3/4 4-15 3/4'4-15 3/4 4-32a 3/4 4-32a 3/4 6-3 3/4,6-3 3/4 8-7 3/4 8-7 3/4 8-10 3/4.8-10 3/4 9-2 3/4 9-2 l

3/4 9-7 3/4 9-7 B3/4 1-1 B3/4 1-1 l

B3/4 1-2 B3/4 1-2 B3/4 2-1 B3/4 2-1 B3/4 3-3 B3/4 3-3 B3/4 3-4 B3/4 3-4 B3/4 4-2

-B3/4:4-2 B3/4 6-2 B3/4 6-2 i

B3/4 7-1 B3/4 7-1 B3/4 7-2 B3/4 7-2

)

B3/4 7-2a B3/4 7-3 B3/4 7-3 1

ATTACHMENT TO LICENSE AMENDMENT NO.'158 (Continued)-

FACILITY OPERATING LICENSE NO. DPR-61 DOCKET NO. 50-213 Replace the following pages of the Appendix. A Technical Specifications with-the enclosed pages. The revised pages are-identified by amendment number and contain vertical lines indicating the areas of change.

Remove Insert-f B3/4 7-6 B3/4 7 B3/4 8-1 B3/4 8-l'

-B3/4 9-2 B3/4 9-2 B3/4 10-1 B3/4 10-1 B3/4 11-1 B3/4 11-1 6-1 6-1 6-5 6-5 6-6 6-6 6-7 6-7 6-8 6-8

+

6-12 6-12 6-20 6-20 6-21 6-21 l

t i

INDEX 1

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS PXaf SECTION 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE j

Leakage Detection Systems...............................

3/4 4-29 Operati on al Le akage.....................................

3/4 4-31 3/4 4-33 3/4.4.7 CHEMISTRY...............................................

TABLE 3.4-1 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS...............

3/4 4-34 TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS i

SURVEILLANCE REQUIREMENTS............................

3/4 4-35 l

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

3/4 4-36 I

FIGURE 3.4-2 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC' ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER l

WITH THE REACTOR COOLANT SPECIFIC ACTIVITY GREATER l

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 Reactor Coolant System..................................

3/4 4-39 l

FIGURE 3.4-3 CONNECTICUT YANKEE LIMIT CURVE FOR HYDROSTATIC AND LEAK TESTING APPLICABLE FOR 22.0 EFFECTIVE FULL POWER YEARS................'.........................

3/4 4-41 FIGURE 3.4-4 CONNECTICUT YANKEE REACTOR COOLANT SYSTEM HEATUP LIMITATIONS FOR 22.0 EFFECTIVE FULL POWER YEARS.....,

3/4 4-42 FISURE 3.4-5 CONNECTICUT YANKEE REACTOR COOLANT SYSTEM C00LdOWN LIMITATIONS FOR 22.0 EFFECTIVE FULL POWER YEARS.....

3/44-43 Pressurizer...........................t..............

3/4 4-45 Low Temperature Overpressure Protectiion 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.

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

3/4 4-51 l

HADDAM NECK VIII Amendment No. JJJ, Jyg, 158 0104

i DDEX q

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS-E8E j

SECTION i

l I

3/4.5 EMERGENCY CORE COOLING SYSTEMS 1

3/4.5.1 ECCS SUBSYSTEMS -.Tavg GREATER THAN OR EQUAL 5TO 350*F...-

3/4 5-1 TABLE 4.5-1 SAFETY INJECTION ACTUATED AUTOMATIC VALVES...........

3/4.5-6 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 1 3/4.5.4 pH CONTROL SYSTEM.......................................

3/4 5-10 J

l 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Containment Integrity...................................

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

3/4.6-2 Containment ~ Air Locks...................................

3/4I6-4 I n te rnal Pre s s u re........................................

3/4 6-6

]

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

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

3/4.6 Containment Ventilation System..........................

3/4 6-9 I

3/4.6.2 CONTAINMENT AIR RECIRCULATION SYSTEM....................

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

3/4 6-12 3/4.7 PLANTS SYSTEMS 2

I l-3/4.7.1 TURBINE CYCLE Safety Va1ves...........................................

3/4.7-1 l

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

3/4 7 Auxiliary Feedwater System..............................

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

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

3/4.7-5 HADDAM NECK IX AmendmentNo.-15,J58 0104 l

3 i

-l

l l

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS t

r E8EE SECTION i

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 l

3/4.7.4 SNUBBERS................................................

3/4 7-10 TABLE 4.7-2 SNUBBER VISUAL INSPECTION INTERVAL...................

3/4 7-10a l

f 3/4.7.5 SEALED SOURCE CONTAMINATION.............................

3/4 7 '

l 3/4.7.6 FIRE SUPPRESSION SYSTEMS Fi re Water Supply /Distributi on System...................

3/4 7-16 i

Spray and/or Sprinkler Systems..........................

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

3/4 7-21 2

H al o n Sy s t ems...........................................

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...................................

3/47-28 3/4.7.8 FLAMMABLE LIQUIDS CONTR0L............-...................

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

3/47-31 TABLE 3.7-6 FEEDWATER ISOLATION VALVES...........................

3/47-32 3/4.7.10 EXTERNAL FLOOD PROTECTION........................~e....'.'.

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

3/4 7-34 x

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 HADDAM NECK X

Amendment No. JEJ, JJJ,158 0104

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS EAEE SECTION 3/4.8.3 ONSITE POWER DISTRIBUTION Operating...............................................

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

3/4 8-14 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRAT I ON.....................................

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

3/4 9-2 3/4.9.3 DECAY TIME..............................................

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

3/4 9-4 i

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

3/4 9-5 3/4.9.6 MAN I PULATOR C RAN E.......................................

3/4 9-6 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE BUILDING..............

3/4 9-7 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION High Water Leve1........................................

3/4 9-8 i

Low Water Leve1.........................................

3/4 9-9 l

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

3/4 9-10 3/4 9.10 WATER LEVEL - REACTOR VESSEL............................

3/4 9-11 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 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWNMARGIN........................j................

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

l 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Concentration...........................................

3/4 11-1 Dose, Liquids...........................................

3/4 11-2 HADDAM NECK XI Amendment No. JJJ, JJ7,158 0104

i INDEX BASES SECTION fAq[

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

B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTR0L........................................

B 3/4 1-1 3/4.1.2 BORATION SYSTEMS........................................

B 3/4 1-2 f

l 3/4.1.3 MOVABLE CONTROL ASSEMBLIES..............................

B 3/4 1-3 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL 0FFSET............................................

B 3/4 2-1 l

3/4.2.2 LINEAR HEAT GENERATION RATE.............................

B 3/4 2-1 3/4.2.3 NUCLEARENTHALPYRISEHOTCHANNELFACTORFfH............

B 3/4 2-1 3/4.2.4 QUADRANT POWER TILT RATI0...............................

B 3/4 2-1 l

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

B 3/4 2-2 l

3/4.3 INSTRUMENTATION 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.......................t....'.'.

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

B 3/4 3-4 j

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

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

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

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

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

B 3/4 4-3 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE.........................

B 3/4 4-5 l

HADDAM NECK XIII Amendment No. JJJ,158 eras

i.

INDEX l

BASES l

I PAGE SECTION 3/4.4.7 CHEMISTRY..............................................

B 3/4 4-7 j

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

B 3/4 4-7 3/4.4.9 PRESSURE / TEMPERATURE LIMITS............................

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

B 3/4 4 ;

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

B 3/4 4-12 B 3/4 4-13'_

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

3/4.5 EMERGENCY CORE COOLING SYSTEMS

)

1 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 3/4.6 CONTAINMENT SYSTEMJ l

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

B 3/4 6-1 3/4.6.2 CONTAINMENT AIR RECIRCULATION SYSTEM....................

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

B 3/4 6-3 l

3/4.7 PLANT SYSTEMS

=

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

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

B 3/4 7-2 i

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

B 3/4 7-2 3/4.7.4 SNUBBERS..........................,......................

B 3/4 7-3 3/4.7.5 SEALED SOURCE CONTAMINATION.............................

B 3/4 7-4 3/4.7.6 FIRE SUPPRESSION SYSTEMS................................

B 3/4 7-4~

3/4.7.7 FIRE RATED ASSEMBLIES...................................

B 3/4 7-5 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 l

3/4.7.11 PRIMARY AUXILIARY BUILDING AIR CLEANUP SYSTEM...........

B 3/4 7-6 HADDAM NECK XIV Amendment No. R5,158 0105

4 i

INDEX BASES fAGI l

4 SECTiON 4

a l

3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1, 3/4.8.2 and 3/4.8.3, A.C. SOURCES, D.C. SOURCES, ONSITE

.ll B 3/4 8-1 POWER DISTRIBUTION........................................

i 4

3/4.9 REFUELING OPERATIONS i

B 3/4 9-1 3/4.9.1 BORON CONCENTRATION.....................................

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

B 3/4 9-1 4

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

{

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

B 3/4 9-1 B 3/4 9-1 3/4.9.5 COMMUNICATIONS..........................................

4 B 3/4 9-2 3/4.9.6 MAN I PU LATOR C RANE.......................................

4 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE BUILDING.,............

B 3/4 9-2 l

3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION...........

B 3/4 9-2 j

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 B 3/4 9-3 P00L....................................................

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 B 3/4 10-1 l

3/4.10.2 PHYSICS TESTS...........................................

3/4.10.3 POSITION INDICATION SYSTEM - SHUT 00WN...................

B 3/4 10-1 3/4.10.4 POSITION INDICATION SYSTEM - OPERATING..................

B 3/4 10-1 3/4.11 RADIOACTIVE EFFLUENTS B 3/4 11-1 3/4.11.1 LIQUID EFFLUENTS........................................

B.3/4'11-2 3/4.11.2 GASEOUS EFFLUENTS.......................................

B 3/4 11-3 3/4.11.3 TOTAL D0SE..............................................

HADDAM NECK XV Amendment No. JJE, JJ7,158 esos

v _a.

l M

i i

ADMINISTRATIVE CONTROLS l

i l

SECTION PAGE 6.0 ADMINISTRATIVE CONTROLS I

6.1 RESPONSIBILITY..........................................

6-1 i

b 4

6.2 ORGANIZATION

)

6.2.1 DNSITE AND OFFSITE ORGANIZATIONS........................

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

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

6-3 i

i 6.3 FACILITY STAFF 0VALIFICATIONS...........................

6-4

/

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

6-5 c

6.5 REVIEW AND AUDIT 6.5.1 PLANT OPERATIONS REVIEW COMMITTEE (P0RC)................

6-5 l

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

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

6-5 i

Alternates..............................................

6-6 i

Meeting Frequency....................................

6-6 j

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

6-6 Responsibilities........................................-

6-6 t

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

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

6-7 6.5.2 NUCLEAR REVIEW BOARD (NRB)...............................

6-7 I

Q u al i fi c at i o n s..........................................

6-7 Composition.............................................

6-8 Consultants.............................................

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

6-8 Quorum..................................................

6-8 Review..................................................

6-9 HADDAM NECK XVIII Amendment No. JJ),158 0106 1

l l

l

~i INDEX ADMINISTRATIVE CONTROLS PKaf SECTION

~6-9 Audits..................................................

6-10 Authority...............................................

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

6.6 REPORTABLE EVENT ACTI0N.................................

6-11 6.7 SAFETY LIMIT VIOLATION..................................

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

6-11 6.9 REPORTING REOUIREMENTS 6.9.1 Routine Reports.........................................

6-13_

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

6-13 An n u al Re p o r t s.......................................... _

Annual Radiological Environmental Operating. Report......

6-14 Semiannual Radioactive Effluent Release Report..........

6-15 i

Monthly Operating Reports................................

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

6-15 S pe c i al Re p o rt s.........................................

6-17 6.10 RECORD RETENTION........................................

6-17

~

6.11 RADIATION PROTECTION PR0 GRAM............................

6-18 1

1 6.12 HIGH RADIATION AREA.....................................

6-19 6.13 RADIOLOGICAL EFFLUENT MONITORING AND OFFSITE DOSE CALCULATION MANUAL (REM 0DCM)............................

6-20 6.14 RADIDACTIVE WASTE TREATMENT.............................

6-20 6.15 SYSTEMS INEGRITY........................................

6-21 6.16 PASS / SAMPLING AND ANALYSIS OF PLANT EFFLUENTS...........

6-21 HADDAM NECK XIX Amendment No. JJJ, JJJ,158 0106

' 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 noted below:

Note 1)

Normally-closed isolation valves SS-SOV-150A, SS-50V-1508, SS-50V-150C, SS-50V-1500, SS-50V-151A, SS-SOV-151B, SS-SOV-151C, and SS-SOV-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)

Normally-closed manual isolation valves SI-V-863A, B, C, and D, SA-V-413, and SS-V-999A may be opened for periodic surveillance and containment boundary (vent and drain) manual valves may be 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 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. JJJ, JJE,158 0107

POWER DISTRIBUTION LIMITS SURVEILLANCE REOUIREMENTS (Continued) 4.2.2.2.2 Measured values of core power peaking factcrs used in determining LHGRs shall include the following allowances:

a.

Normal power peaking * **,

b.

Flux peaking augmentation factors (Power Spike)*,

l c.

Measurement uncertainty of 1.05, d.

Statistical density factor of 1.012, e.

Engineering factor of 1.02, f.

Stack shortening / thermal expansion factor of 1.007, and g.

Power level uncertainty of 1.02.

1 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. JZE,158 naos

TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION g

-o E

MINIMUM TOTAL NO.

CHANNELS CHANNELS APPLICABLE

=

y 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 4

3.

Wide Range, Neutron Flux, 4

2 3

2,3*,4*,5*

2, 4 High Start Up Rate Trip 1(a)'

6f 4.

Pressurizer Pressure-Variable, low 4

2 3

5.

Pressurizer Pressure--High 3

2 2-1, 2 6#

6.

Pressurizer Water Level--High 3

21 2.

1, 2, 3***'

'68 l

b 7.

Reactor Coolant Flow - Low 2/ loop 2/ loop.

I Ib) 6#

a.

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

b.-

Above P-7 and 3/ loop 2/ loop

' loop I

k Below P-8 in each

.in any two

~ 15. Each E.

operating

-loops **

operating i

l loop loop a

. g.

i-w.

i

. ~.

TABLE 3.3-1 (Continued)

Sig REACTOR TRIP SYSTEM INSTRUMENTATION

!3 -

MINIMUM 3;

TOTAL NO.

CHANNELS CHANNELS APPLICABLE p; -FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE-MODES ACTION 7t 8.

Steam Flow-High 4 (1/ steam line) 2 1/ steam line 1, 2 9f 9.

Steam Generator Water 1/SG level 1/SG level 1/SG level 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 I ")

8 I

10. Undervoltage ' Reactor ~

2 (1/ bus) 1 2 (1/ bus)-

R Coolant Pumps l

a td

11. Safety Injection 2

1 2.

1, 2 -

12.

6 9

j n

o

?

5; m

I

TABLE 3.3-1 (Continued)

TABLE NOTATION With the Reactor Trip System breakers in the closed position and the 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.

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

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 ACTION 1:

With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 2:

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

a.

The inoperable channel is placed in the tripped condition within 6

hours, l

b.

The Minimum Channels OPERABLE requirement is met;

however, the inoperable channel may be bypassed for up to_4 hours for surveillance testing of other channels per Specification 4,.3.1.1.

ACTION 3:

a.

With less than the Minimum Number of' Channels OPERABLE, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months determine by observation of the associated permissive annunciator window (s) that the interlock is in its required state for the existing plant condition or apply Specification 3.0.3.

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. J#,158 0110

i i

TABLE 3.3-1 (Continued)

ACTION STATEMENTS (Continued)

ACTION 10:

i

~

f With the number of OPERABLE channels'one less than the Minimum Channels i

OPERABLE requirement for Modes 3, 4, 5, restore the inoperable. channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open the Reactor Trip System breakers within the next hour.

ACTION'll:

With the number of OPERABLE channels one.less than the' Minimum Channelst OPERABLE' requirement, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

l' I

ACTION 12:-

With the number of OPERABLE channels one less than the minimum channels l

OPERABLE requirements,.be in at least HOT STANDBY.within 6_ hours; however,-

)

one channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for. surveillance testing per-Specification 4.3.1.1, provided the other channel'is OPERABLE.

k i

j t

I i

i a

s i

i

't HADDAM NECK 3/43-7

' Amendment No. JU, 158 j

sits L

TABLE 4.3-1 (Continued) j TABLE NOTATIONS l

,1 4

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 subcritical.

THERMAL POWER greater than 10% of RATED THERMAL POWER.

(a)

(b) THERMAL POWER greater than or equal to 74% of RATED THERMAL POWER.

(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 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 conditions. When this exception is used, a heat balance calculation will continue to be performed oc a daily basis to determine core power, and the power range channels will be verified daily to be 9% below the selected overpower trip setpoint.

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

The TRIP ACTUATING DEVICE OPERATIONAL TEST shall independently verify (4) 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.

t s

HADDAM NECK 3/4 3-12 Amendment No. JU, JM,158 0112

.l 1

TABLE 3.3-2 fContinued) l TABLE NOTATIONS

Trip function may be bypassed in this MODE when RCS pressure is less than j

l 1800 psig.

J

- The channel (s) associated with the protective functions derived from the 1

out-of-service reactor coolant loop shall be placed in the tripped mode.

q l

{

(a) THERMAL POWER greater than 10% of RATED THERMAL POWER.

(b) For Surveillance Testing, at most only one train may. be taken out of.

3 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 4.3-2) the l

minimum channels OPERABLE.may be less than those specified in Table 3.3 for items 3.1.a, 3.a.2, and 6.a.

ACTION STATEMENTS l

With the number of. OPERABLE channels one less.than the ACTION 20 Minimum Channels OPERABLE requirement, be in at least HOT STANDBY.within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD. SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

With the number ' of OPERABLE channels one less than the

'I ACTION 21 Minimum-Channels OPERABLE requirement, startup and/or power operation may proceed until performance of the next required ANALOG CHANNEL OPERATIONAL TEST.provided the l

inoperable channel is placed in. the tripped condition within I hour.

i With a channel associated with an operating loop inoperable,

'1 ACTION 22 restore the inoperable channel to OPERABLE shtus within 4 I

hours or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

With the number : of OPERABLE channels one.less than the ACTION 23 Minimum Channels OPERABLE requirement,. restore-the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following.30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

I HADDAM NECK-3/4 3-17 Amendment No. J U, JJJ,158 0113 l

~-

...-. u.

o 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:

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

l I hour, and j

b.

The Minimum Channels OPERABLE requirement is met; however, one additional channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .for surveillance testing of other cha.nnels per Specification 4.3.2.1.

l ACTION 25 Not used.

l With the number of OPERABLE channels one less than the ACTION 26 Minimum Channels OPERABLE requirement, restore the. inoperable channel to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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 ~ 0PERABLE requirement, restore the inoperable channel to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or_ place the DC powered hydraulic pump in service. Otherwise, be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

l E

l l

i l

p l

HADDAM NECK 3/4 3-18

- Amendment No. JU, J#,158 0114

TABLE 3.3-4 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS E-MINIMUM CHANNELS CHANNELS APPLICABLE ALARM / TRIP zy FUNCTIONAL UNIT TO TRIP / ALARM OPERABLE MODES SETPOINT ACTION 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 w1 ACTION 30 -

Must satisfy the ACTION requirement for Specification 3.4.6.1.

t Y2 a

it a.

1

?

4 TABLE 3.3-8 (Continued) l FIRE DETECTION SYSTEMS Minimus Number Minimus Number Smoke Detectors Heat Detectors OPERABLE / Detectors OPERABLE / Detectors Location Available Available 4/4

17. Turbine building mezzanine under l

l generator (T-1F) j J

l

~6/6 l

IB. Turbine building cranewell deluge (T-lC)

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

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

2 l

i l

l HADDAM NECK 3/4 3-43 Amendment No. JJJ,158 0116

1*

t i

f.

REACTOR COOLANT SYSTEM i

ISOLATED LOOP LIMITING CONDITION FOR OPERATION 1

3.4.1.5 The RCS loop stop valves of an isolated loop

shall be shut and:

either:

a.

The power removed from the valve operators, or b.

The boron concentration of the isolated loop shall be maintained greater than or equal to the boron concentration of the operating i

loops.

l APPLICABILITY: MODES 1 and 2 ACTION:

d I

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

f a.

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

Increase the boron concentration of the_ isolated loop to within l

the limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , or c.

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

I SURVEILLANCE REQUIREMENTS

l 4.4.1.5.1 At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , if required, verify that power 'is I

l removed from the valve operators.

(

4.4.1.5.2 At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , if required, verify that the boron l

concentration of an isolated loop is greater than _or. equal to the boron concentration of the operating loops.

i.

i l

A loop is considered to be~ isolated whe'n the hot and cold leg stop valves are both closed.

Three-loop operation is not allowed for Cycle '17. _

l_

l HADDAM NECK-3/4 4-9 Amendment No. JJ), JAS, _158 l

-osar

q i

REACTOR COOLANT SYSTEM r

l IDLED LOOP c

LINITING CONDITION FOR OPERATION' i

i i

3.4.1.8 The cold leg loop stop valve of an idled loop

shall. be. shut and.

9 either: #

.j The power removed from the' valve operator, or a.

b.

The boron concentration' of.the. idled loop shall be maintained greater than or equal to. the boron concentration ' of the. operating ~

loops..

APPLICABILITY: ' MODES I and 2.

ll ACTION:

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

a.

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

b.

Increase the boron concentration of the. idled loop to wit' in the h

limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , or c.

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

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

q SURVEILLANCE REQUIRENENTS J

4.4.1.8.1 At least once per l2' hours,. 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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , if required,, verify that the. boron concentration of an idled loop is greater -than or equal to the boron concentration.of the operating loops.

3 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.

i HADDAM NECK 3/4 4-12 Amendment No. JJ), Jpp,158 0119

...--., ~,

-,,,,..u-.

1 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.

The boron concentration of the idled loop shall be maintained 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 refueling boron concentration of Specification 3.9.1.

APPLICABILITY:

MODES 3, 4, 5, and 6 l

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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , or Be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD c.

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

SURVEILLANCE REOUIREMENTS 4.4.1.9.1 At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , 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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , 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 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.

i HADDAM NECK 3/4 4-13 Amendment No..J #,158-eus

REACTOR COOLANT SYSTEM IDLED LOOP STARTUP LIMITING CONDITION FOR OPERATION 3.4.1.11 A reactor coolant loop shall remain idled until:

The temperature at the cold leg of the idled loop is within 20*F a.

of the highest cold leg temperature of the operating loop (s),*

b.

The boron concentration of the idled loops is 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 refueling boron concentration of Specification 3.9.1.

APPLICABILITY:

MODES 3, 4, 5, and 6 l

ACTION:

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

SURVEILLANCE 4.4.1.11.1 The idled loop cold leg temperature shall be determined to be 1

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 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 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.

x 4.4.1.11.3 At least once per refueling outage the stop valve / temperature 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.

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.

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. JJJ,158 0120

1 p

i 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation due to flow through the valve.

j

)

The provisions of Specification 4.0.4 are not applicable for entry j

into MODE 3 or 4 for Specification h, above.

j i.

In addition to surveillance requirement 4.4.6.2.1.g at least once por 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 i

exception of RHR suction piping).. The test shall be conducted at a l

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 i

2.

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

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

accident conditions.

t 3.

Tiping 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.

t

- Except for those portions of the HPSI, Charging and' RHR suittion piping which

~

are not testable at accident pressure during normal operation, as defined below.

HPSI 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 sucti,ons.=

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#, JM, JM,158 erzi

~

..-...c.

i U

~

SURVEILLANCE REQUIREMENTS (Continued) b.

If anyL 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-J to meet' 0.75 La, a Type A testc shall be performed at least every 18 months until-two consecutive Type'A tests meet 0.75 La at which j

time the above test schedule may be resumed or a corrective action 4

i plan may be prepared and submitted to the NRC that provides.an acceptable alternative contingent on NRC approval.<-

The accuracy of each Type A test shall be verified using th'e

'l c.

i relationship:

)

(LTM + l - 0.25 L,) s L I IlTM + A + 0.2 O,)

o c

o where:

L is_.the percent measured containment leakage per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at TM l-pressure P

2 t

-l L,

is the percent superimposed' leakage,.

L is the percent leakage obtained from the! supplemental-l c

test result, and

\\

L, is replaced with L for reduced pressure tests.

t d.

Type B and C. tests shall'be. conducted.at intervals :no greater than 1

24 months and at.a pressure not less than Pa, 39.6 psig, using l

halogen gas detection, soap-bubble, pressure decay, or other methods'of equivalent sensitivity, except for tests involving:

I 1)

Air locks, and 2)

Purge supply and exhaust isolation valves with resilient material seals.

e.

Air locks shall be tested and demonstrated OPERABLE by-the requirements of Specification 4.6.1.3; f

f.

Purge supply and exhaust isolation valves with resilient material seals shall be tested and demonstrated OPERABLE by the requirements of Specification 4.g.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.

HADDAM NECK 3/4 6-3 Amendment No. J#, J,4J,158 0122

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

-.,: s

ELECTRICAL POWER SYSTEMS SHUTDOWN 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.

ACTION:

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.

SURVEILLANCE RE0VIREMENTS s

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 (exceptsfor Specification 4.8.1.1.2a.5, 4.8.1.1.2b, 4.8.1.1.2f), and 4.8.1.1.3.

HADDAM NECK 3/4 8-7 Amendment No. J2E, JfE, 158 0123

TABLE 4.8-2 l

BATTERY SURVEILLANCE REOUIREMENTS II)

Quarterly (2)

Weekly PARAMETER LIMITS FOR EACH LIMITS FOR EACH ALLOWABLEI3)

DESIGNATED PILOT CONNECTED CELL VALUE FOR EACH CELL CONNECTED CELL Electrolyte

> Minimum level

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

plates, 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) 11.190 Not more than Gravity 0.020 below the average of all connected cells average of all Average of all connected cells connecg cells

>1.200

>1.195 P

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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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 i

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.

1 i

i l

HADDAM NECK 3/4 8-10 Amendment No. J#, 158 0124

REFUELING OPERATIONS 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 (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:

With one of the above required monitors inop'erable or not operating, a.

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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , b.

An ANALOG CHANNEL OPERATIONAL TEST within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to the initial start of CORE ALTERATIONS', and c.

An ANALOG CHANNEL OPERATIONAL TEST at least once per 7 days.

HADDAM NECK 3/4 9-2 Amendment No. J U, Jf7 158 0125 i

I

{

~

r i

REFUELING OPERATIONS

?

l

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

l LIMITING CONDITION FOR OPERATION F

i 3.9.7 Loads in. excess of -1650 pounds shall be prohibited from travel 'over ;

fuel assemblies in the storage pool.

l i

APPLICABILITY: With fuel assemblies in the storage pool.

ACTION:

With the requirements of. the above specification not1 satisfied, a.

place the crane load'in a safe condition.

b.

The provisions of Specification 3.0.3 are not applicable..

.l SURVEILLANCE REQUIREMENTS 4.9.7 Administrative controls that prevent --the travel : of-loads ~ in excess i

of 1650 pounds over fuel assemblies shall be in place prior to lifting a -

load in excess of 1650 pounds.

-l

)

2 J

j I

~HADDAM NECK 3/4 9-7 Amendment'No. J U,158-esas m,

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

m 4

~

1 3/4.1 REACTIVITY CONTROL SYSTEMS 1

BASES 3/4.1.1 BORATION CONTROL l

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 1

associated with postulated accident conditions are controllable within acceptable limits, and_ (3) the reactor will be maintained sufficiently j

suberitical to preclude inadvertent criticality.in the shutdown condition.

y b

SHUTDOWN MARGIN requirements vary throughout core life as a function of. fuel The most restrictive ~

depletion, RCS boron concentration, and RCS =T conditionduringMODES1,.2and3occursat.eM9a.f-cyc1 11fe (EOL), and is associated with a postulated steam line break accident and resulting RCS q

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-o 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? n MODES 4 and 5'is 1

i 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.

.I 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 lihdting 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 spe'cified conditions is acceptable. Correction. factors shall account for fuel and moderator temperature and boron concentration.

HADDAM NECK B3/4 1-1 Amendment No.: JAl, Jfp,158 0127

=-

' REACTIVITY CONTROL SYSTEMS BASES MODERATOR TEMPERATURE COEFFICIENT (Continued)

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 lowly due principally to the reduction in RCS boron concentration associated with fuel burnup.

3/4.1.1.6 MINIMUM TEMPERATURE FOR CRITICALITY l

This specification ensures that the reactor will not be made critical with i

i the Reactor Coolant System average temperature less than 525'F. This limitation is required to ensure: (1) the moderator temperature coefficient is 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 supply 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 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 EOL from full power equilibrium xenon conditions, and the minimum required volume of 12,000 gallons of,14,000-ppm barated water from the boric acid tank meets this requirement.

With the RCS temperature below 200*F, one boration system is acceptable i

without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional rest.rictions prohibiting CORE ALTERATIONS and positive reactivity changes in the' event the single boration system becomes inoperable.

I HADDAM NECK B3/41-2 Amendment No. JD, Jg,158 0127 I

1 j

j 3/4.2 POWER DISTRIBUTION LIMITS y

BASES I

1 l

The specifications of this section provide assurance of fuel integrity _

during Condition I (Nomal Operation) and II (Incidents of Moderate Frequen-j cy) events by:. (1) maintaining _the minimus DNBR in the core greater than or.

]

equal to 1.30 during norma 1' operation and in short-tem transients, and (2) limiting the fission gas release, fuel pellet temperature, and cladding l

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 j

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.-

j 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.

j NUCLEARENTHALPYRISEHOTCHANNELFACTORFf3, 3/4.2.3 The limit on the NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (FAH) ensures that the mirtimum DNBR limit is not exceeded.

N The F is measurable, but will normally only be determined' periodically as speciNedinSpecification4.2.3.1.2and4.2.3.2.2.

This periodic surveill3nce is sufficient to insure that the limits are maintained i

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.

The relaxation of F as a function of THERMAL POWER allows changes 'inithe radial power shape Ur 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. JJJ, JJJ;158 :

aus

=

m O

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

Firedetectorsthatare.usedtoactiateFireSuppression.Systemsrepresenta more critically important component of,a' plant's Fire Protection' Program l

than detectors that:are installed solely for early fire warning and notifi-cation. Consequently,_ the minimum number of OPERABLE fire detectors must be greater.

The loss of detection capability-for Fire Suppression Systems,' actuated by '

j l

fire detectors, represents a significant degradation of. fire protection for-As-a result, the establishment of a fire watch patro1 must be-1 1

any area.

l 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 areasyis required'to provide detectionJ capability until the-inoperable instrumentation is' restored to'0PERABILITY.-

ll 3/4.3.3.7 RADI0 ACTIVE LIOUID EFFLUENT MONITORING INSTRUMENTATION' The radioactive liquid effluent instrumentation:is provided to monitor and' l

control, as applicable, the. releases of radioactive materialsE n. liquid-i effluents during actual or potential releases'of: liquid effluents. The j

AlarnVTrip Setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters-in the.0DCM to ensure.that l

the alarm / trip will occur prior to exceeding the limits of 10. CFR Part 20 The OPERABILITY and use of this instrumentation is consistent with the-1 requirements of General Design Criteria 60, 63,' and 64 of Appendix ' A to._

-l 10 CFR Part 50.

j l

3/4.3.3.8 RADIDACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION'

)

The radioactive gaseous effluent instrumentation is provided.to. monitor and-control, as applicable, the. releases of radioactive materiaTs in~ gaseous 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 REMODCM 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 -

J tent with the requirements of General-Design' Criteria 60,- 63, and<64 of Appendix A to 10 CFR Part 50.

3/4.3.3.9 BORON DILUTION ALARM The shutdown monitors proride indication of positive' reactivity insertion.

during operation in Modes 3, 4, 5, and 6.

The indication is credited in the Boron Dilution. design basis analysis.

l l

HADDAM NECK B3/4 3-3 Amendment No. J#,_158 0129

-, _ ~.

,m,._

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

I INSTRUMENTATION BASES 3/4.3.4 INTERNAL FLOOD PROTECTION l

The liquid level instrumentation is provided to monitor liquid levels ~in the areas of potential flooding caused by local pipe ruptures. The system ensures 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-able instrumentation is restored to OPERABILITY.

I l

h l

l l

i 2

i

(

i i

I l

HADDAM NECK B3/4 3-4 Amendment No. JJ), 158 1

0129

{

. a.-.

l l

[

o l

REACTOR COOLANT SYSTEM

]

BASES l

l 3/4.4.1 REACTOR COOLANT SYSTEM LOOPS AND COOLANT CIRCULATION (Continued) l The restrictions on starting an RCP with one or more RCS cold legs less thani 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.s 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 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 boron concen -

tration in an isolated / idled loop prior to opening thetstop 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.

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 240,000-1bs. per. hour at 2,485 psig as 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 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 capability 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 abova 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_

Reactor trip _until the first Reactor Trip System Trip Setpoint is reached-L (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 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 ASME Boiler and Pressure Code.

HADDAM NECK B3/4 4-2 Amendment'No. JU,158 0131 u

m

_,n. _.2

__ _,_ _,., _,u - _, ; a

.m._. -_

l' i

1 i

CONTAlletENT SYSTEMS j

BASES

. 3/4.6.1.5 AIR TEMPERATURE fContinued) 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 CONTAINMENT VESSEL-STRUCTURAL INTEGRITY

'[

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 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 The 42-inch containment purge _ supply and exhaust isolation v'alves and the 1

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. LTo l

provide assurance _ that_ these containment valves cannot' be inadvertently l

opened, the valves are locked closed in accordance with' Standard Review Plan j

6.2.4 which includes mechanical devi_ces 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 the containment following a loss-of-coolant. accident....The containment air particulate monitoring system provides the primaryt means of purging because it provides adequate' purge flow to prevent _ an explosive lm'ixture 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.

l exhaust penetration inside the reactor containment so that the containment.

I 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.

HADDAM NECK-

.B3/4 6-2 AmendmentNo.j#,158 euo

.a.

._ _.a.;_. _

j e

t 3/4.7 PLANT SYSTDtS j

BASES 1

3/4.7.1 TURBINE CYCLE j

3/4.7.1.1 SAFETY VALVES

-l The OPERABILITY of the main steam line Code safety valves ensures that the j

Secondary Coolant System pressure will be limited to below 1105,-(1100 j

3 psia), of Its design pressure of:1000 psia during,the most. severe anticipated. system operational transient.-iThe maximum relieving capacity is' 1

associated with a Turbine tri from 1005 RATED THERMAL POWER coincident with i

an assumed loss of condenser at 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 Pressure Vessel-14 Code, 1971 Edition. The design total relieving capacity for all valves on.

all of the steam lines is g,504,000 lbs/hr which is 120% of the total secondary steam flow of 7,872,000 lbs/hr at 1005 RATED THERMAL POWER.

l 3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM l

The OPERABILITY of the Auxiliary Feedwater System ensures that the Reactor l

Coolant System can be cooled down-to less than 350*F from normal' operating' j

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

~

i Each steam turbine-driven auxiliary feedwater pmp has a capacity sufficient i

to ensure adequate delivery of feedwater. flow to remove decay heat and.

i reduce the' Reactor Coolant System temperature;to less-than 350*F within the j

Residual Heat Removal System operating range.:- With one auxiliary feedwater; 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 j

two steam turbine-driven feedwater pumps inoperable,"at 1 east one pump must

{

be restored to OPERABLE with 24. hours 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 4

be restored to OPERABLE within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months from the time'of initial loss of.the.

I first pump or be in HOT STANDBY;in the next six hours and HOT SHUTDOWN l

within the following six hours.

q The auxiliary feedwater (AFW) system's design basis fequires AFW to be automatically initiated and to be independent of any AC electrical power.

source for at least two hours. The AFW pump / turbine governor's DC powered-1 hydraulic pump, controls, and DC power supply are required to be 0PERABLE i

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 domineralized water storage tank (DWST) and primary water storage tank (PWST) with the minimum water volume ensures thatt sufficient water is available to maintain the RCS at HDT STANDBY conditions for 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months with steam discharge to the atmosphere concurrent with total.

HADDAM NECK B3/4 7-1 Amendment No. JJJ, JJJ, JJJ,158

. 0133

-, -. =...,

-l,.,

I 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 initiated 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 i

j is adequate for decay heat removal for a period of at least 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . Within this period, decay heat removal demands are reduced to approximately 150 gps. Makeup water is available during this period from the PWST which contains a minimum volume of 80,000 gallons. The PWST transfer pumps can l

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 Coolan' System specific activity ensure that the resultant offsite radiation dose will be limited to a small fraction of j

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.

l 3/4.7.1.5 MAIN STEAM LINE TRIP VALVES The OPERABILITY of the main steam line trip valves ensures that no more than i

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,Regoirements are consistent with the assumptions used in the safety analyses.

3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION The limitation on steam generator pressure and temperature ensures that the pressure-inducedstressesinthesteamgeneratorsdonotexceedthgmaximum 5

allowable fracture toughness stress limits. Thg limitations of 70 F and 200 psigarebasedonasteamgeneratorRTNDTof10Fandaresufffcientto 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.

4 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. JJJ, Jfy 158 j

B133

l PLANT SYSTEMS BASES 3/4.7.3 SERVICE WATER SYSTEM (Continued) 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 pgs associated with each diesel generator, the motor operated isolation valve that isolates -

nonessential turbine building service water loads, the air operated.

isolation valve that isolates nonessential primary auxiliary building service water loads, the motor operated isolation valve that aligns service water flow to the RHR heat exchanger, and the i

t f

i 9

E t

i 1

I HADDAM NECK B3/4 7-2a Amendment No. JJJ, JA7,15S 0133

l 1

q 1'

PLANT SYSTEMS BASES 3/4.7.3 SERVICE WATER SYSTEM (Continued) i i

safety-related piping and components.

j Each service water header contains an in-line filter with a bypass line f

containing a motor operated isolation valve that supplies the containment'-

l air recirculation unit coolers. The bypass line allows continued plant i

operation should the 'in-line filter become clogged. Each bypass line and its isolation valve are sized such that only one bypass line is needed to pass the required service water flow. To satisfy single failure

[

assumptions, an operable in-line filter bypass isolation valve is required

i in each service water header due to possible common mode failure of the' i

in-line filters and at least one in-line filter must be in service whenever the service water system is required to be OPERABLE.,The service' water _

headers may be tied together by an open service water header cross-connect in the intake structure.

3/4.7.4 SNUBBERS All snubbers are required to be OPERABLE to ensera that the structurni integrity of the reactor coolant system and all other safety-related systems-is maintained during and following a seismic or other event initiating dynamic loads. Snubbers excluded from this inspection program are those-installed on nonsafety-related systems and then only if their failure, or i

failure of the system on which they are installed, would have no adverse effect on any safety-related system.

j The visual inspection frequency is based upon maintaining a constant level i

of snubber protection to systems..Therefore, the required inspection interval varies inversely'with the observed snubber failures and is deter-mined by the number of inoperable snubbers found during an inspection.

Inspections performed before that interval has elapsed may'be used as a new reference point to determine the next inspection.. However,' the results of.

such early inspections performed before the original required time interval has elapsed (nominal time less 25%) may not be used Jo lengthen the required inspection interval. Any inspection whose results require a shorter inspec-tion interval will' override the previous schedule.

When the cause of the rejection of a snubber is clearly established 'and remedied for that snubber and for any other snubbers that may be generically

)

susceptible and verified by inservice functional testing, that snubber may be exempted from being counted as inoperable. Generically. susceptible I

l snubbers are those which are of a specific make or model and have the same l-design' features directly related to rejection of.the snubber by visual inspection or are similarly located or exposed to the same. environmental conditions, such as temperature, radiation, and vibration.

l HADDAM NECK B3/4 7-3 Amendment No. JJJ, JJ7.158

'l l

3133 l

I.

j PLANT SYSTEMS

=)

f BASES i

3/4.7.9 FEEDWATER ISOLATION VALVES The accident analysis for a main steam line break assumes that the main.

l feedwater isolation valves will close on a containment isolation actuat%n signal (CIAS). Also, the closure of these valves based on a CIAS is credited i

in detemining the Pressure / Temperature limits for the purpose of l

environmental qualification. 'The.feedwater isolation valves act as a backup-.

l to the feedwater regulation valves in the event a feedwater regulation valve I

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.

i Normal i

protection to safety-related equipment to El. 30 feet :nean sea level.

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 AOP 3.2-24.

3/4.7.11 PRIMARY AUXILIARY BUILDING AIR CLEANUP SYSTEM

.l PAB Air Cleanup System consists of two exhaust fans, two prefilters, a HEPA-HECA filter assembly, and interconnecting ductwork.-

l Air cleanup is accomplished using one exhaust fan, one prefilter, the l

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 belpg ' purged) the purge discharge l

would be directed through the Primary Auxiliary B'uilding charcoal filters.

Credit is again taken for these filters in r, educing the radiological consequences.

1 4

HADDAM NECK B3/4 7-6 Amendment No. JJJ,158 0134 l

s A

3/4.8 ELECTRICAL POWER SYSTEMS BASES 3/4.8.1. 3/4.8.2 AND 3/4.8.3 A. C. SOURCES. D. C. SOURCES. ONSITE 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 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 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.

i I

power sources and associated distribution systems OPERABLE during accident conditions coincident with an assumed loss-of-offsite power and single 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 l

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 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 1977; and 1.137, " Fuel-Oil Systems for Standby Diesel Generators", Revision 1, October 1979, and guidance given in Generic Letter 84-15.

l HADDAM NECK B3/4 8-1 Amendment No. JU,158 ons

l e

~

REFUELING OPERATIONS-BASES I

3/4.9.6 MANIPULATOR CRANE I

The OPERABILITY requirements for the manipulator cranes ensure that:

(1) l 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 i

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.

~

l 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE BUILDING

[

f 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 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 j

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 j

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.

1

'i 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 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 release of radioactive material from the containment atmosphere to the

~

environment.

HADDAM NECK B 3/4 9-2

. Amendment No. J S 158 i

0136 i

,.,a

f 3/4.10 SPECIAL TEST EXCEPTIONS BASES l

3/4.10.1 SHUTDOWN MARGIN I

This Special Test Exception provides that a minimum amount of control rod worth is immediately available for reactivity control when tests are per-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 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 chaPUteristics of the core and related instrumentation can be verified.

In order for various characteristics to be accurately measured, it is at times necessary to 6

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 l

normally be allowed by Specification 3.1.3.6 which in turn may cause the RCS i

3 Y9.o fall slightly below the minimum temperature of Specification T

t 8

14.

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, position indication system calibration, and slave cycler testing. The exception is required during rod drop testing since the data necessary to determine the rod drop time are derived from the induced voltage 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.

The exception is also necessary during position indication system calibration and slave cycler testing, since these activities require withdrawal of control rods in order to perform calibration and ensure proper i

with the highest worth bank fully withdrawn ensures shutdown mar h s 0.94 operation of the control rod drive cyclers.

Since maintaining K i

requirements are met, position indication is not required, and the reactivity requirements for rod drop testing are bounding.

3/4.10.4 POSITION INDICATION SYSTEM - OPERATUM i

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

HADDAM NECK B3/4 10-1 Amendment No. J#, JJ/,157.15 0103 1

~

s i

c 3/4.11 RADIOACTIVE EFFLUENTS i

BASES l

i 3/4.11.1 LIOUID EFFLUENTS 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 This limitation provides additional assurance that B, Table II, Column 2.

i the levels of radioactive materials in bodies of water outside the site will

-l result in exposures within:

(1) the Section II.A design objectives of,

Appendix 1,10 CFR Part 50, to a MEMBER OF THE PUBLIC, and (2) the limits of 10 CFR 20.106(c) to the population. The. concentration limit for dissolved l

or entrained noble gases is based upon the assumption that Xe-135 is the l

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.

l I

3/4.11.1.2 DOSE. LIOUIDS This specification is provided to implement the requirements of Sections II.A. III.A, and IV.A of Appendix I, 10 CFR Part.50. The Limiting Condition for Operation implements the guides set forth in Section II.A of Appendix I.

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 3

kept "as low as is reasonably achievable". The dose calculation methodology and parameters in the REM 0DCM 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 i

for calculating the doses due to the actual: release rates of radioactive j

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 1," 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.

i l

HADDAM NECK B3/4 11-1 Amendment No. UE,158 0138 i

-ea

6.0 ADMINISTRATIVE CONTROLS

{

r 6.1 RESPONSIBILITY l

6.1.1 The Vice President - Haddam Neck.shall be responsibl'e for overall l

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

t 6.2 ORGANIZATION i

6.2.1 ONSITE AND OFFSITE ORGANIZATIONS f

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

.j nuclear power plant.

Lines of authority, responsibility, and communication shall' be l

a.

established and defined for the highest management levels through t

l intermediate levels to and including all operating organization.

i positions. These relationships shall be' documented and updated, l

as appropriate, in the' form of organization. charts, functional.

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.

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-plant.

c.

The Executive Vice President-Nuclear, shall have corporate" i

responsibility for overall plant nuclear safety and shall take any

-l measures needed to ensure acceptable performance ~ of the staff in i

i 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 1

sufficient organizational freedom to ensure their independence from operating pressures.

j 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 Operator shall be in the control room; HADDAM NECK 6-1 Amendment No. JJJ, JJJ,158 0139

ADMINISTRATIVE CONTROLS l

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 accidents.

6.4 TRAINING l

l 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 responsibility for the implementation of the Training Program.

j 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 intent of Section 27 of the NFPA Code-1975, except for: Fire Brigade training sessions which shall be held at least quarterly.

i I

6.5 REVIEW AND AUDIT-6.5.1 PLANT OPERATIONS' REVIEW COMMITTEE (PORC) j FUNCTION I

t 6.5.1.1 The PORC shall function.to advise the_ Vice President-Haddam Neck i

on all matters related to nuclear safety.

COMPOSITION 6.5.1.2 The PORC shall be composed of the:

Chairperson:

Vict President - Haddam Neck:

l Member:

Nuclear Unit Director Member:

Operations Manager Member:

Maintenance Manager Member:

Instrument and Control Manager

[

Member:

- Reactor Engineer-l Member:

Engineering Manager.

Member:

Nuclear Services Director Member:

Plant Quality Services Supervisor Member:

Chemistry Manager Member:

Health Physics Manager Member:

Security Manager HADDAM NECK 6-5 Amendment No. JJE, JJJ, JEE,158 t.

0140 o

l

l q

E f

ADMINISTRATIVE CONTROLS Il-t ALTERNATES r

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 time.

l MEETING FRE00ENCY 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.

l t

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

designated alternate and four members including alternates.

i l

l RESPONSIBILITIES i

l 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; i

i b.

Review of all proposed tests and experiments that affect nuclear

.j safety; Review of all proposed changes to the Technical Specifications;.

c.

d.

Review of all proposed changes or modifications to plant systems-1 or equipment that affect nuclear safety; e.

Investigation of all violations 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 HADDAM NECK 6-6 Amendment No.- JEE,. JEE,158 0140

l ADMINISTRATIVE CONTROLS 1

i i.

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

Review of the Emergency. Plan and implementing procedures and. shall su' omit recommended changes to the Chairperson of the Nuclear Review Board; AUTHORITY 6.5.1.7 The PORC shall:

Report to and be advisory to the Vice President - Haddam Neck on l

a.

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 l

the PORC and the Vice President - Haddam Neck; however, the Vice President - Haddam Neck shall have responsibility for resolution l

l 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 NUCLEAR REVIEW BOARD (NRB)

OVALIFICATIONS 6.5.2.1 The minimum qualifications of NRB members are as follows:

I 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 i

i 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

i experience.

HADDAM NECK 6-7 Amendment No. J/E, JEE,158 ouo

3 ADMINISTRATIVE CONTROLS b.

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

1.

Nuclear power plant operations i

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 i

COMPOSITION 6.5.2.2 The NRB shall consist of no less than eight, nor more than eleven l

members including the Chairperson and the Vice President - Haddam Neck.

The 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 FRE00ENCY 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 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 (.hairperson or a designated alternate. No more than a minority of the quorum shall have line responsibility for operation of the facility.

l l

l I

HADDAM NECK 6-8 Amendment No. JEE, JEE,158 0140

i ADMINISTRATIVE CONTROLS The applicable procedures recommended 4 Appendix A of Regulatory a.

Guide 1.33, Revision 2, February 1976; b.

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

Fire Protection Program implementation.

c.

d.

Quality controls for effluent monitoring, using the guidance in l

Regulatory Guide 1.21 Rev.1, June 1974.

l RADIOLOGICAL EFFLUENT MONITORING AND OFFSITE DOSE CALCULATION e.

l 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 reviewed by the PORC and shall be approved by the Vice President - Haddam Neck prior to implementation and reviewed periodically as set forth in each l

document or in administrative procedures.

6.8.3 Temporary changes to procedures of Specification 6.8.1 may be made l

provided:

The intent of the original procedure is not altered; l

a.

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

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 REM 0DCM.

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 approved by appropriate supervision.

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.

HADDAM NECK 6-12 Amendment No. JEE, JEE, 158 0141

ADMINISTRATIVE CONTROLS L

6.13 RADIOLOGICAL EFFLUENT MONITORING AND OFFSITE DOSE CALCULATION MANUAL (REM 0DCM)

Section I,

Radiological Effluents Monitoring Manual, shall outline the sampling and analysis programs to determine the concentration of radioactive released offsite as well as dose commitments to individuals in materials those exposure pathways and for those radionuclides released as a result of facility operation.

It shall also specify operating guidelines for RADICACTIVE 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 performed 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 radioactive effluent control required by 10 CFR 20.106, 40 CFR Part 190,10 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 approval of the Vice President - Haddam Neck.

l Shall be submitted to the Commission in the form of a complete, legible c.

of the entire REMM or ODCM, as appropriate, as a part of or copy 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 RADIOACTIVE 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 PADI0 ACTIVE 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 i

shipping and burial ground requirements.

HADDAM NECK 6-20 Amendment No. JEE, JEE,158 0142

_. - ~..

o-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:

5 Provisions establishing 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.

'6.16 PASS / SAMPLING AND ANALYSIS OF PLANT EFFLUENTSL j

The licensee shall implement and maintain-a: program which will ensure the capability to obtain and analyze reactor. coolant, radioactiveciodines and particulates in plant gaseous effluents, and containment. atmosphere samples under' accident conditions..This program lshall. include the following; a.

Training of personnel b.

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

c.

i i

'i

)

1 i

i 1

HADDAM NECK

'6-21 Amendment No.-J7E, JEE,158:

0143 la

ML20044H047

Text

Attachment:

Navigation menu

Search