Text

Amend 49 to License DPR-3,revising Tech Specs to Correct Typographical & Editorial Errors & Improving Clarity & Consistency of Several Tech Spec Requirements
	ML20031C387
Person / Time
Site:	Yankee Rowe
Issue date:	05/30/1978
From:	Ziemann D; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20031C382	List: ML20031C381; ML20031C387; ML20031C391; ML20031C393;
References
	NUDOCS 8110070105
	Download: ML20031C387 (96)
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4 UNITED STATES y

NUCLEAR REGULATORY COMMISSION g

WASHINGTON. D. C. 20555

\\.,...../

YANKEE ATOMIC ELECTRIC COMPANY DOCKET NO. 50-29 YANKEE NUCLEAR OWER STATION (YANKEE-ROWE)

AMENDMENT TO FACILITY OPERATING LICEN E Amendment Nc.49 License No. DPR-3 1.

The Nuclear Regulatory Commission (%e Commission) has found that:

A.

The aoplication for amendment by Yankee Atomic Electric Company (the licensee) dated March 23, 1977, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility s -11 operate in conformity with the application, the provision.s af the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the puolic, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

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

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

8110070105 780530' PDR ADOCK 05000029 P

PDR

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

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- 2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragrapn 2.C.(2) of Facility License No. DPR-3 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 49, 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 its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION h

(,J mu. c < N W

Dennis L. Ziemann, Chief Operating Reactors Branch #2 Division of Operating Reactors

Attachment:

Changes to the Technical Specifications Date of Issuance: May 30,1978 l

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ATTACHMENT TO LICENSE AMENDMENT NO. 49 FACILITY LICENSE NO. DPR-3 DOCKET NO. 50-29 Revise Appendix A Technical Specifications by removing the following pages and inserting the enclosed pages. The revised pages are identified by the captioned amendment number and contain vertical lines indicating the area of change. Overleaf pages are included for document completeness.

Remove Insert VII VII XII XII 3/4 1 3/4 1-6 3/4 1 3/4 1-6 3/4 1-18 3/4 1-18 3/4 1-19 3/4'l 19 3/4 1 3/4 1-22 3/4 1 3/4 1-22 3/4 3-2 3/4 3-2 3/4 3-8 3/4 3-8 3/4 3-10 3/4 3-10 3/4 3-1E - 3/4 3-16 3/4 3 3// 3-16 3/4 3-18 3/4 3-18 3/4 3-19 3/4 3-19 3/4 3-23 3/4 3-23 3/4 3 3/4 3-26 3/4 3 3/4 3-26 3/4 4 3/4 4-2 3/4 4 3/4 4-2 3/4 4-4 3/4 4-4 3/44-7 3/4 4-7 3/4 4-23 3/4 4-23 3/4 4-25 3/4 4-25 3/4 5-6 3/4 5-6 3/4 5-9 3/4 5-9 3/4 5-11 3/4 5-11 3/4 6 3/4 6-2 3/4 6 3/4 6 3/4 6-2 3/4 6-4 3/4 6-4 3/4 6-8 3/4 6-8 3/4 6 3/4 5-10 3/4 6 3/4 6-10 3/4 6-14 3/4 6-14 3/4 7-9 3/4 7-9 3/4 7-17 3/4 7-17 3/4 7-21 3/4 7-21 3/4 7-24 3/4 7-24 3/4 8-6 3/4 8-6 3/4 8-7 3/4 8-7 3/4 10-3 3/4 10-3 8 3/4 4-1 B 3/4 4-1 B 3/4 4-8 B 3/4 4-8 8 3/4 6-3 B 3/4 6-3 5-4 5-4 6-6 6-6 6-7 6-7 6-10 6-10 6-21 6-21

f INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS PAGE SECTI0h 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES 3/4 8-1 0perating..............................................

~

3/4 8-S Shutdown..........................................

3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS A.C. Di stributi on - Operati ng..........................

3/4 8-6 A. C. Di s trib u ti on - Sh9tdown...........................

3/4 8-8 D. C. Di s tri bu ti on - O pera ti ng..........................

3/4 8-9 D. C. Di stri buti on - Shutdown...........................

3/4 8-11 3/4.9 REFUELING OPERATIONS 3/4 9-1 3/4.9.1 REACTIVITY.............................................

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

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

3/4 9-5 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS......................

3/4 9-6 3/4.9.5 COMMUNICATION5.........................................

3/4.9.6 SHIELD TANK CAVITY MANIPULATOR CRANE OPERABILITY.......

3/4 9-7 3/4 9-8 3/4.9.7 CRANE TRAV EL - EP ENT FU EL P IT..........................

3/4 9-9 3/4.9.8 COOLANT CIRCULATIO!4....................................

3/4 9-10 4

3/4.9.9 CONTAINMENT PURGE FAN SHUT 00WN SYSTEM..................

3/4 9-11 3/4.9.10 W ATER L EV EL-REACTOR V ESSEL.............................

3/4 9-12 3/4.9.11 SPENT FUEL PIT WATER LEVEL.............................

YANKEE-ROWE VII Amendment No.49 m

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s INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.9.12 SPENT FUEL Pli BUILDING IS0LATION......................

3/4 9-13 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN........................................

3/4 10-1 3/4.10.2 CONTROL ROD OPERABILITY AND INSERTION LIMITS...........

3/4 10-2 3/4.10.3 PRESSURE / TEMPERATURE LIMITATION-REACTOR CRITICALITY....

3/4 10-3 3/4.10.4 PHYSICS TESTS..........................................

3/4 10-4 YANKEE-ROWE VIII

t INDEX BASES SECTION PAGE 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT..............................

B 3/4 6-1 3/4.6.2 CONTAINMENT ISOLATION VALVES............................

B 3/4 6-3 3/4.6.3 COMBUSTIBLE GAS CONTR0L.................................

B 3/4 6-3 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE..........................................

B 3/4 7-1 3/4.7.2 STEAM GENERATOR PXLSSURE/TEMERATURE LIMITATION..........

B 3/4 7-4 3/4.7.3 PRIMARY PUMP SEAL WATER SYSTEM..................

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

B 3/4 7-4 3/4.7.5 CONTROL ROOM VENTILATION SYSTEM EMERGENCY SHUTDOWN......

B 3/4 7-4 3/4.7.6 SEALED SOURCE CONTAMINATION.............................

B 3/4 7-4 3/4.7.7 WASTE EFFLUENTS.........................................

B 3/4 7-5 3/4.7.8 ENVIRONMENTAL MONITORING................................

B 3/4 7-6 3/4.7.9 SHOCK SUPPRESSORS (SNUBBERS)............................

B 3/4 7-6 3/4.7.10 FIRE SUPPRESSION SYSTEMS................................

B 3/4 7-7 3/4.7.11 PENETRATION FIRE BARRIERS..........................

B 3/4 7-7 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES....................

B 3/4 8-1 3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS.........

D 3/4 8-1 3/4.9 REFUELING OPERATIONS 3/4.9.1 REACTIVITY..............................................

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

B 3/4 9-1 YANKEE-ROWE XI Amendment No. 47

3 INDEX BASES l

SECTION_

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

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

B 3/4 9-1 3/4.9.5 COM1UNICATIONS..........................................

B 3/4.9-1 3/4.9.6 SHIELD TANK CAVITY MANIPULATOR CRANE OPERABILITY....... B 3/4 9-2 l

3/4.9.7 CRANE TRAVEL - SPENT FUEL PIT...........................B 3/4 9-2 3/4.9.8 COOLANT CIRCULATION....................................

B 3/4 9-2 3/4.9.9 CONTAINMENT PURGE FAN SHUTDOWN SYSTEM..................

B 3/4 9-2 3/4.9.10 and 3/4.9.11 WATER LEVEL-REACTOR VESSEL AND SPENT FUEL P1T....................................................

B 3/4 9-2 3/4.9.12 SPENT FUEL PIT BUILDING IS0LATION......................

B 3/4 9-3 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN........................................

B 3/4 10-1 3/4.10.2 CONTROL R00 OPERABILITY AND INSERTION LIMITS...........

B 3/4 10-1 3/4.10.3 PRESSURE / TEMPERATURE LIMITATIONS-REACTOR CRITICALITY............................................

B 3/4 10-1 3/4.10.4 PHYSICS TESTS..........................................

B 3/4 10-1 YANKEE-ROWE XII Amendment No. Af 49 t

l REACTIVITY CONTROL SYSTEMS BORON DILUTION LIMITING CONDITION FOR OPERATION 3.1.1. 3 Main Coolant System baron concentration shall not be reduced unless:

The flow rate of main coolant to the reactor pressure vessel is a.

1 950 gpm; b.

The maximum reactivity insert {on rate due to baron concentra-tion reduction is < l.5 X 10- ak/k per second; and c.

Main coolant temperature is t 250*F.

APPLICABILITY All MODES.

ACTION:

With the flow rate of main coolant to the reactor pressure vessel a.

< 950 gpm, immediately suspend all operations involving a reduction in boron concentration of the Main Coolant System.

With the maximum reactivity insertion rate due to main coolant b.

baron concentration reducticn in excess of the limit, immediately suspend boron concentration reduction, and verify the required SHUTDOWN MARGIN within one hour.

With the main coolant temperature < 250*F, immediately suspend c.

boron concentration reduction and verify the required SHUTDOWN MARGIN within one hour.

SURVEILLANCE REQUIREMENTS 4.1.1.3.1 The flow rate of main coolant to the reactor pressure vessel l

shall be determined to be > 950 gpm within one hour prior to the start of and at least once per hour during a reduction in the Main Coolant System baron concentration by either:

YANKEE-ROWE 3/41-5 Amendment No.49

REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) a.

Verifying at least one main coolant pump is in operation l

or b.

Verifying that the shutdown cooling system is in opera-l tion and supplying > 950 gpm to the reactor pressure vessel.

4.1.1.3.2 Isolation valves of ion exchangers capable of reducing main l

coolant baron concentration shall be verified to be locked closed at least once per 31 days except when the ion exchanger is in use for boron removal.

YANKEE-ROWE 3/4 1-6 Amendment No.49

8

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REACTIVITY CONTROL SYSTEMS BORIC ACID MIX TANK GRAVITY FEED CONNECTION - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.8 The boric acid mix tank gravity feed connection in the boron injection flow path required by Specification 3.1.2.3 shall be OPERABLE if the flow path from the boric acid mix tank in Specification 3.1.2.3 is OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With the boric acid mix tank gravity feed connection inoperable, restore the boric acid gravity feed connection to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and borated to a SHUTDOWN MARGIN (all control rods inserted) equivalent to 5% ak/k at 200*F; restore the boric acid gravity feed connection to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.1.2.8 The above required baric acid mix tank gravity feed connection shall be demonstrated OPERABLE:

a.

At least once per 7 days by:

1.

Cycling Each testable power operated valve in the flow path through at least one complete cycle of full travel.

4 2.

Verifying that each valve (manual or power operated) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position, b.

At least once per 18 months, during shutdown, by demonstrating the gravity feed conne<, tion flow to be > 26 gpm.

TANKEE-ROWE 3/4 1-17 l

s N

REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES - REFUELING LIMITING CONDITION FOR OPERATION 3.1.2.9 As a minimum, the boric acid mix tank and associated heat tracing shall be OPERABLE with:

a.

A minimum contained borated water volume of 1500 gallons, equivalent to a tank level _> 3.6 feet, b.

12 to 12.5% by weight boric acid solution, and c.

A minimum solution temperature of 150*F.

APPLICABILITY: MODE 6.

ACTION:

With the boric acid mix tank inoperable, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until the boric acid mix tank is restored to OPERABLE status.

1 SURVEILLANCE REQUIREMENTS 4.1.2.9 The boric acid mix tank shall be der.cnstrated OPERABLE at least once per 7 days by:

Verifying the boric acid concentration of the water, l

I a.

b.

Verifying the contained borated water volume of the tank, and' l

I c.

Verifying the boric acid mix tank solution temperature.

l l

l YANKEE-ROWE 3/4 1-18 l

l Amendment No.49 L

t REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES - SHUTOOWN LIMITING CONDITION FOR OPERATION 3.1.2.10 As a minimum, one of the following borated water sources shall be OPERABLE:

The boric acid mix tank and associated heat tracing with:

a.

1.

A minimum contained borated water volume of 1500 gallons, equivalent to a tank level > 3.6 feet, 2.

12 to 12.5f. by weight boric acid solution, and 3.

A minimum solution temperature of 150*F.

b.

The safety injection tank (SIT) with:

1.

A minimum contained borated water volume of 117,000 gallons, equivalent to a tank level of > 25.5 feet.

2.

A minimum boron concentration of 2200 ppm, and 3.

A minimum solution temperature of 40*F.

APPLICABILITY: MODE 5.

ACTION:

With no barated water scurce OPERABLE, suspend all operations involving positive reactivity changes until at least one barated water source is restored to OPERABLE status.

SURVEILLANCE REQUIREMENTS i

4.1.2.10 The above required borated water source shall be demonstrated OPERABLE:

(

a.

At least once per 7 days by:

1.

Verifying the boron concentration of the safety injec-tion tank water or the boric acid concentration of the

[

boric acid mix tank water.

2.

Verifying the contained borated water volume of the tank, and YANKEE-ROWE 3/4 1-19 l

Amendment tio. 49 l

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,r REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued}

l 3.

Verifying the boric acid mix tank solution temperature 4

when it is the source of borated water.

4 b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the SIT temperature

'when it is the source of borated water and the outside air temperature is <

35'F.

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YANKEE-ROWE 3/4 1-20

REACTIVITY CONTROL SYSTEMS 80 RATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERA 110N 3.1.2.11 Each of the following borated w.ter sources shall be OPERABLE:

a.

The boric acid mix tank and associated heat tracing with:

1.

A minimum contained borated water volume of 1500 gallons, equivalent to a tank level of,3.6 feet, 2.

12 to 12.5% by weight boric acid solution, 3.

A minimum solution temperature of 150*F.

b.

The safety injection tank (SIT) with:

1.

A minimum contained borated water volume of 117,000 gallons of water, equivalent to a tank level of > 25.5 feet, 2.

A minimum boron concentration of 2200 ppm, and 3.

A minimum solution temperature of 40'F.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With either the boric acid mix tank or the safety injection tank inoperable, provided the other required source is OPERABLE, restore the inoperable tank to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at lea' t HOT STAND 8Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and borated to s

a SHUTDOWN MARGIN (all control rods inserted) equivalent to at least 5% ak/k at 200*F; restore the inoperable tank to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the l

i

? ext 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS l

4.1.2.11 Each borated water source shall be demonstrated OPERABLE:

l YANKEE-ROWE 3/ 4 1-21 Amendment No. 49

{

REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREME'lTS (Continued) a.

At least once per 7 days by:

1.

Verifying the boron concentration of the safety injection tank water, and the boric acid concentration of the boric acid mix tank water.

2.

Verifying the contained borated water volume of each water source, and i

3.

Verifying the boric acid mix tank solution temperature.

b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the SIT temperature when the outside air temperature is < 35'F.

I i

YANKEE-ROWE 3/4 1-22 Amendment No. 49 1

e3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTIVE SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the reactor protective system instrumer.tation channels and reactor permissive functions of Table 3.3-1 shall be OPERABLE.

APPLICABILITY: As shown in Table 3.3-1.

ACTION:

As shown in Table 3.3-1.

SURVEILLANCE REQUIREMENTS 4.3.1.1 Each reactor protective system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the OPERATIONAL M0uES and at the frequencies shown in Table 4.3-1.

4.3.1.2 The logic for the Reactor Permissive Circuit shall be demon-strated OPERABLE during the at power CHANNEL FUNCTIONAL TEST of channels affected by permissive circuit operation. The total permissive function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by permissive circuit operation.

YANKEE-ROWE 3/4 3-1

TABLE 3.3-1 REACTOR PROTECTIVE SYSTEM INSTRUMENTATION MINIMUM y

]l; TOTAL NO.

CHANNELS CHANNELS APPLICABLE N

FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION Tg 1.

Manual Reactor Trip 3

1 3

1, 2 and

1 E

2.

Power Range, Neutron Flux and Intermediate Power Range, Neutron Flux 6

2 4

1, 2 and *{j) 2**

3.

Intermediate Range, Neutron Flux, High Startup Rate 2

1 2

1(2), 2 and

3 l

4.

Source Range, Neutron Flux Startup##

2 NA 2

2# and *(5) 4 l

a.

m b.

Shutdown 2

NA 1

3,4,5(5) 5 l

S.

Low Main Coolant Flow (SGAP) 4 2

3 1(3) 6**

l 6.

Low Main Coolant Flow (MC Pump Current) a.

System A 4

2 3

1 7**

b.

System B 4

2 3

1 7**

g I4)

E 7.

Low Pressurizer Pressure 1

1 1

1, 2 8

8 I4I A

8.

Low Main Coolant System Pressure 1

1 1

1, 2 8

I4) 9.

High Pressurizer Water Level 1

1 1

1, 2 8

10.

Low Steam Generator Water Level 4

2 3

1(3) 6**

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4 TABLE 3.3-1 (Continued)

ACTION STATEMENTS (Continued)

ACTION 7 (Continued) -

b)

The flinimum Channels OPERABLE requirement for each System is met; however, one additional channel in either system 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 Specifi-cation 4.3.1.1.

ACTION 8 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 9 - With the number of channels OPERABLE one less than required by the Minimum Channels Operable requirement, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months with reactor trip breakers open.

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YANKEE-ROWE 3/4 3-7

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TABLE 4.3-1 REACTOR PROTECTIVE SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS i

E CHANNEL MODES IN WHICH E

CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE "4

FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRFD o

III E

1.

Manual Reactor Trip NA NA S/U NA 2.

Power Range, Neutron Flux and Intermediate Power Range.

Neutron Flux S

D(2),Q(5)

M 1, 2 and l

3.

Intermediate Range, Neutron Flux, S

R(5)

M 1, 2 and l

High Startup Rate i.

4.

Source Range, Neutron Flux S

R(5)

S/U 2,3,4,5 and l

U S.

Low Main Ccolant flow (SGaP)

S R(4)

M(3) j i'

m 6.

Low Main Coolant Flow, Systems A and B (MC Pump Current)

S R

H 1

I4I M(3) 1, 2 7.

Low Pressurizer Pressure S

R I4)

I3) 8.

Low Main Coolant System Pressure S

R M

1, 2 I4)

M(3) 1, 2

[

9.

High Pressurizer Water Level S

R I4) g 10.

Low Steam Generator Water Level S

R M

1 5

?

j

'e Y

k

TABLE 4.3-1 (Continued)

REACTOR PROTECTIVE SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS z

mm b

CHANNEL MODES IN WilCH E

CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED

11. Turbine Trip flA NA S/U(I) 1 III llA NA S/U 1

12. Generator Trip S/U(I) 1, 2 and *

13. Reactor Trip Breaker IIA NA II)

14. Automatic Trip Logic flA NA S/U 1, 2 and M

Y.

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TABLE 4.3-1 (Continued)

NOTATION With the reactor trip system breakers closed and the control rod drive system capable of rod withdrawal.

If not performed in the previous 7 days.

(1)

I Heat balance only, above 15% of RATED THERMAL POWER, at least (2) 3 times per week with a maximum time interval of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

When shutdown longer than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , if not performed in the (3) previous 31 days.

(4)

Known pressure applied to sensor.

Neutron detectors may be excluded from CHANNEL CALIBRATION.

l (5)

.J 1

YANKEE-ROWE 3/4 3-10 Amendment No. 49

l

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TABLE 4.3-2 ENGINEERED SAFEGUARDS SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS A

CHANNEL MODES IN WHICH T

E CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE E

FUNCTIONAL ENIT CHECK CAllBRATION TEST REQUIRED 1.

SAFETY INJECTION a.

Actuation Channel 1 S

NA M(1) 1, 2, 3#

l

1) RPS Low Main Coolant i

Pressure Channel -

S R(3)

M(2) 1, 2, 3#

2) High Containment Pressure Sensor S

R(3)

M(3) 1, 2, 3#

3) Manual Initiat h N.A.

N.A.

R 1, 2, 3, 4, 5*

l M

b.

Actuation Channel 2 S

N.A.

M(1) 1, 2, 3#

l f

1) Low Pressurizer Pressure Sensor S

R(3)

M(3) 1, 2, 2#

l

2) High Containment Pressure Sensor S

R(3)

M(3) 1, 2, 3#

3) Manual Initiation H.A.

N.A.

R I, 2, 3, 4, 5*

l 2.

CONTAINMENT ISOLATION a.

Manual Initiation N.A.

N.A.

R 1, 2, 3, 4, 5* I b.

Actuation Channel S

N.A.

M(4) 1,2,3,4,5* I

1) High Containment Pressure Sensor S

R(3)

M(3) 1, 2, 3, 4, 5* l i

4 e

TABLE 4.3-2 (Continued)

TABLE NOTATION (1) When shutdown with main coolant pressure < 1000 psig, if not performed within the previous 31 days.

l (2) When shutdown longer than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , if not perfomed in the previous 31 days.

(3) The test shall include exercising the sensor by applying either a vacuum or pressure to the appropriate side of the sensor.

(4) When in COLD SHUTDOWN with main coolant pressure < 300 psig, if not performed within the previous 31 days.

flot required in this MODE with main coolant pressure < 300 PSIG.

tiot required in this MODE with main coolant pressure < 1700 PSIG.

6 YANKEE-ROWE 3/4 3-16 Amendment No. 49

_._-___-_----_---------m--

INSTRUMENTATION 3/4.3.3 MONITORING INSTRUMENTATION RADIATION HONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION i

3.3.3.1 The radiation monitoring instrumentation channels shown in Table 3.3-4 shall be OPERABLE with their alarm setpoints within the specified limits.

APPLICABILITY: As shown in Table 3.3-4.

ACTION:

a.

With a radiation monitoring channel alarm setpoint exceeding the Alarm Setpoint shown in Table 3.3-4, adjust the setpoint to within the limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or declare the channel inoperable.

b.

With one or more radiation monitoring channels inoperable, 4

take the ACTION shown in Table 3.3-4.

The provisions of Specifications 3.0.3 and 3.0.4 are not c.

applicable.

I SURVEILLANCE REQUIREMENTS l

l 4.3.3.1 Each radiation monitoring instrumentation channel shall be l

demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the OPERATIONAL MODES and at the frequencies shown in Table 4.3-3.

l l

l YANKEE-ROWE 3/4 3-17

TABLE 3.3-4 RADIATION MONITORING INSTRUMENTATION EE MINIMUM si CHANNELS APPLICABLE ALARM MEASUREMENT

'7' INSTRUMENT OPERABLE MODES SETPOINT RANGE ACTION 8

Ni 1.

AREA MONITORS a.

Spent Fuel Pit Area

1) Fuel Manipulator 1

< 5 mr/hr or 2 x 0.5 - 50 mr 20 Gamma Guard Eackground, which-ever is greater b.

Containment

1) Fuel Manipulator I

< 10 mr/hr or 2 x 0 - 1000 mr 21 Gamma Guard background, which-ever is greater w

1:

2.

PROCESS MONITORS w

1.

a.

Containment 4

1) Main Coolant System 1

1,2,3,&4

< 30 CPS above 10 - 10 CPS 22 Leakage Air Particu-background late Monitor b.

Radioactive Gaseous Waste Monitor

< 2000 cpm or 10-10fcpmor

1) Loop Seal 1

23 2-Discharge Monitor F x background 10 - 10 cpm j

whichever is greater o.

2) Primary Vent Stack 5

Monitor 6

a) Particulate Monitor 1 At all times

< 900 cpm greater 10 - 10 cpm 24 se than background 8

4 s

M

TABLE 3.3-4 (Continued) 5 k

MINIMUM CHANNELS APPLICABLE-ALARM MEASUREMENT E

INSTRUMENT OPERABLE MODES SETPOINT RANGE ACTION E

6 b) Iodine Monitor 1

At all times

< 700 cpm greater 10 - 10 cpm 24 than background c) Noble Gas 6

Monitor 1

At all times 1 3500 cpm greater 10 - 10 cpm 24 than background j

i c.

Radioactive Liquid Monitors II}

3 1, 2, 3 & 4

< 80 cps or 2 x 1 x 10 cps 23

1) Steam Generator I

Blowdown Monitor Eackground, wnich-2 ever is greater i

3.

ACCIDENT-EMERGENCY MONITORS.

a.

High Level Radiation

]

Monitor 1

At all times 1 5 R/hr 0.01 - 1000 R/hr 23 k=

a

?

i

6 TABLE 3.3-4 (Continued) s TABLE NOTATION j

When handling irradiated fuel, control rods, or sources.

With radioactive effluent in the waste gas surge drum.

(1)

Per steam generator in a non-isolated loop.

ACTION STATEMENTS ACTION 20 -

With the number of channels OPERABLE less than required by the flinimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.9.12.

ACTION 21 -

With the number of channels OPERABLE less than required by the tiinimum Channels OPERABLE requirement, suspend all operations involving CORE ALTERATIONS.

ACTION 22 -

With the number of channels OPERABLE less than required by the l4inimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.4.5.1.

ACTION 23 -

With the number of channels OPERABLE less than required by the 141nimum Channels OPERABLE requirement, provide an OPERABLE temporary continuous monitor within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

ACTION 24 -

With the number of channels OPERABLE less than require'd by the Itinimum Channels OPERABLE requirement, suspend all planned releases and releas9s from the evaporator to the atmosphere through the primary vent stack.

YANKEE-ROWE 3/4 3-20

INSTRUMENTATION INCORE DETECTION SYSTEM LIMITING CONDITION FOR OPERATION 3.3.3.2 The incere detection system shall be OPERABLE 'with:

a.

At least 75% of the neutron detector thimbles OPERABLE, b.

A minimum of 2 OPERABLE neutron detector thimbles per core quadrant, and c.

Sufficient OPERABLE movable neutron detectors, drive, and readout equipment to map these thimbles.

d.

At least ten OPERABLE racisi position thermocouples with an OPERABLE thermocouple in at least one of the two calculated hottest instrumented fuel tssemblies.

APPLICABILITY: When the incore detection system is used for core po e r distribution measurements.

ACTION:

With the incore detection system inoperable, do not use the system for the above applicable monitoring or calibration functions. The provisions l

of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEIL'_ANCE REOUIREMENTS 4.3.3.2 The incore neutron detectors shall be demonstrated OPERABLE by

' nonnalizing each detector output to be und wichin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to its use for core power distribution measurements.

YANKEE-ROWE 3/4 3-23

a INSTRUMENTATION i

METEOROLOGICAL INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.3 The niteorological monitoring instrumentation channels shown in Table 3.3-5 shall be OPERABLE.

APPLICABILITY: At all times.

ACTION:

With one or mrre required meteorological monitoring channels inoperable for more thaq 7 days, prepare and submit a Special Report to the Cannission pursuant tc Specification 6.9.6 within the next 10 days outlining the cause of '.he malfunction and the plans for restoring the channel (s) to OPERABl!

atus. The provisions of Specifications 3.0.3 and 3.0.4 are not a,,

.able.

SURVEILLANCE REQUIREMENTS 4.3.3.3 Each of the above meteorological monitoring instrumentation channels shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-4.

I l

l l

( YANKEE-ROWE 3/4 3-24 l

t

P TABLE 3.3-5 METEOROLOGICAL MONITORING INSTRUMENTATION MINIMUM INSTRUMENT LOCATION OPERABLE 1.

WIND SPEED a.

Nomin 1 Elev. 195 feet 1

4 b.

Nominal Elev. 33 feet 1

2.

WINO DIRECTION a.

Nominal Elev.195 feet 1

4 b.

Nominal Elev. 33 feet 1

3.

AIR TEMPERATURE - DELTA ~

a.

Nominal Elev 195-33 feet 1

i i

f.

YANKEE-ROWE 3/4 3-25 Amendment No. 49 v

TABLE 4.3-4 s

METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLAtlCE REQUIREMENTS.

CHANNEL CHANNEL INSTRUMENT CHECK CALIBRATION 1.

WIND SPEED a.

Nominal Elev.195 feet D

SA b.

Nominal Elev. 33 feet D

SA 2.

WIND DIRECTION a.

Nominal Elev.195 feet D

SA b.

Nominal Elev. 33 feet D

SA l

3.

AIR TEMPERATURE - DLETA T a.

Nominal Elev.195-33 feet D

SA YANKEE-ROWE 3/4 3-26 Amendment No. 49

3/4.4 MAIN COOLANT SYSTEM 3/4.4.1 MAIN COOLANT LOOPS NORMAL OPERATION LIMITING CONDITION FOR OPERATION 3.4.1.1 All main coolant loops shall be in operation with all loop isolation valves, in a non-isolated loop, OPERABLE with power removed per Specification 4.5.2.b.4.

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

ACTION:

a.

MODE 1: With one main coolant loop and associated pump not in operation, STARTUP and/or continued POWER OPERATION may proceed provided THERMAL POWER is restricted to less than 75% of RATED THERMAL POWER and the following Reactor Protective System instrumentation channels associated with the loop not in operation, are placed in their tripped condition within 1 hour:

1. Low Main Coolant Flow (Steam generator AP)

2. Low Main Coolant Flow (Main coolant pump current)

3. Low Steam Generator Water Level b.

MODES 2,3,4, and 5:

1.0, operation may proceed provided at least With K 1.

two maYcc>olant loops and associated pumps are in operation.

ff < 1.0, operation !::oy proceed provided at least 2.

With K one maTn coolant loop and associated pump is in opera-tion or the Shutdown Cooling System is in operation, except when there is a substantial amount of decay heat in the reactor fuel, then at least two main coolant loops shall be tied to the reactor vessel (main coolant pumps may be Off), or the Shutdown Cooling System shall be in operation.

With the reactor vessel and connecting pres-MODES 3,4, and 5:

c.

surizer system isolated from the heat removal system by closing the loop isolation valve (s), leak testing may be performed provided that the coolant temperature in the reactor vessel does not in-crease at a rate exceeding 50* per hour, the maximum temperature increase during the test period does not exceed 100*F, and pres-surizer pressure does not exceed 2485 psig.

YANKEE-ROWE 3/4 4-1 Amendment No. 49

f MAIN COOLANT SYSTEM LIMITING CONDITION FOR OPERATION (Continued) d.

MODES 1,2,3, and 4: With one or more main coolant system loop isolation valves in non-isolated loop inoperable, restore the inoperable valve (s) to operable status prior to exceeding 200*F T,yg.

The provisions of Specifications 3.0.3 and 3.0.4 are not e.

applicable.

SURVEILLANCE REQUIREMENTS 4.4.1.1.1 In MODE 1, with one reactor coolant loop and associated pump not in operation, at least once per 31 days determine that the applicable Reactor Protection System instrumentation channels specified in the ACTION statements above have been placed in their tripped conditions.

4. 4.1.1. 2 In MODES 2,3,4 or 5, determine that the steam generators associated with the main coolant loops required to be in operation are l

capable of decay heat removal by verifying at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months that:

The Main Coolant System is closed and pressurized to > 100 psi a.

above saturation pressure.

b.

The Main Coolant System loop cold and hot leg stop valves are fully open, with the bypass valve closed.

The steam generator water level is above the top of the tube bundle.

c.

An inventory of over 85,000 gallons of primary grade feedwater is d.

available.

A boiler feed pump is OPERABLE.

e.

Verify that the Shutdown Cooling System isolation valves are 4.4.1.1.3 locked closed within one hour prior to increasing Main Coolant System pressure above 300 PSIG.

At least once per 18 months, during shutdown, demonstrate

4. 4.1.1. 4 main coolant loop isolation valve operability by cycling each valve through at least one complete cycle of full travel from the control room.

YANKEE-ROWE 3/4 4-2 Amendment No. 49

MAIN COOLANT SYSTEM ISOLATED LOOP LIMITING CONDITION FOR OPERATION 3.4.1.2 The boron concentration of an isolated loop shall be maintained greater than or equal to the boron concentration of the operating loops.

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

ACTION,:

With the requirements of the above specification not satisfied, do not open the isolated loop's stop valves; either increase the boron concen-tration of the isolated 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 STANDBY within the next 6 '9urs with the unisolated portion of the Main Coolant System borated to a SHUTDOWN MARGIN (all rods inserted) equivalent to at least 5% ak/k at 200*F.

SURVEILLANCE REOUIREMENTS 4.4.1.2.1 The baron concentration of an isolated loop shall be determined to be greater than or equal to the baron concentration of the operating loops at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . With Main Coolant System loop isolation valve controls removed from service and mechanically prevented from t

l operation by lock and key, the frequency of determination that the baron concep+=ation of the isolated loop is greater than or equal to the baron stion of the operating loops may be reduced to 3 times per week conc l

with u inaximum of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months between analyses.

4.4.1.2.2 An isolated loop shall be determined to be borated to at least 5% AK/K at 200*F before loop temperature is reduced > 30*F below the highest cold leg temperature of the operating loops. With Main Coolant System loop isolation valve controls, main steam line isolation valves and main coolant pump controls removed from service and mechanically prevented from operation by lock and key, the baron concentration of the isolated loop may be reduced to that existing in the Main Coolant System at that time.

l YANKEE-ROWE 3/4 4-3 I

MAIN COOLANT SYSTEM ISOLATED LOOP STARTUP LIMITING CONDITION FOR OPERATION 3.4.1.3 A main coolant loop shall remain isolated until:

a.

The temperature of the cold leg of the isolated loop is within 30*F of the highest cold leg temperature of the operating l

loops.

b.

The boron concentration of the isolated loop is not less than the main coolant system boron concentration, and c.

The reactor is subcritical by at least 1 percent ak/k APPLICABILITY: All MODES.

ACTION:

With the requirements of the above specification not satisfied, suspend startup of the isolated loop.

SURVEILLANCE REQUIREMENTS 4.4.1.3.1 The isolated loop cold leg temperature shall be determined to be within 20*F of the highest cold leg temperature of the operating loops within 30 minutes prior to opening the cold leg stop valve.

4.4.1.3.2 The isolated loop boron concentration shall be determined to be not less than the Main Coolant System boron concentration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to opening the cold leg stop valve.

4.4.1.3.3 The reactor shall be determined to be subcritical by at least 1 percent ak/k within 30 minutes prior to opening the cold leg stop valve.

YANKEE-RCWE 3/4 4-4 Amendment No. 49 l

MAIN COOLANT SYSTEM PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.4 The pressurizer shall be OPERABLE with a steam bubble.

APPLICABILITY: MODES 1 and 2 ACTION:

With the pressurizer inoperable, be in at least HOT STANDBY with the reactor trip breakers open within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.4.4 The pressurizer shall be determined operable:

At least once per 18 months by determining that; a.

1.

Pressurizer circulation spray flow is sufficient to require 500-700 KW per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months pressurizer heater power consumption.

~

j 2.

Sufficient pressurizer surge spray flow exists by opening the spray valve and observing the resultant pressurizer pressure decrease and heater cycling.

b.

At least once per 18 months, and after completion of any maintenance on the solenoid relief valve, by detemining that the solenoid relief valve opens at 2400 + 30 psig and closes at 2350 + 30 psig on a test signal ~from the l

i RPS pressurizer pressure channel.

l I

YANKEE-ROWE 3/4 4-7 Amendment No.49 i

l

MAIN LOOLANT SYSTEM s

3/4.4.5 MAIN COOLANT SYSTEM LEAKAGE-LEAKAGE DETECTION SYSTEMS LIMITING CONDITION POR OPERATION I

3.4.5.1 The following Main Coolant System leakage detection systems shall be OPERABLE:

a.

The containment atmosphere particulate radioactivity monitoring

system, l

b.

The containment drain tank level monitoring system.

I c.

The incore detection system thimble leak alarm system.

APPLICABILITY: MODES 1, 2, 3 and 4.

l ACTION:

a.

Witn the above required radioactivity monitoring leakage detection system inoperable, operation may continue for up to 7 days provided:

1.

Main Coolant System water inventory balance is perfomed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2.

The other above required leakage detection systems are OPERABLE,'

and 3.

Appropriate grab samples are obtained and analyzed at least once per hour:-

otherwise, 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 .

b.

With the containment drain tank level monitoring system inoperable, restore the inoperable system to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 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 .

c.

With the incore detection system thimble leak alam system inoper-able, restore the leak alarm system to OPERABLE status within 7 days or close all thimble isolation valves; restore the leak alam system to OPERABLE status within 31 days or be in at least HOT STANDBY with-in the 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 .

YANKEE-ROWE 3/4 4-J

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00 nezcarrn ruessame en Allowable Pressure-Temperature Limitations During Cooldown for Inservice Test FIGURE 3.4-5 YANKEE-ROWE 3/4 4-24

8 MAIN COOLANT SYSTEM e

PRESSURIZER LIMITING CONDITION FOR OPERATION

~

3.4.8.2 The pressurizer temperature shall be:

a.

Limited to a maximum heatup 'of 100*F in any one hour period, b.

Limited to a maximum cooldown of 200*F in any one hour period, c.

Within 225'F of the Main Coolant System temperature, and d.

Greater than 70*F whenever pressurizer pressure exceeds 500-t-

psig.

~

APPLICABILITY: At all timei.

l_

ACTION:

With the cressurizer temperature outside of any of the above limits.

restore the temperature to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the fracture toughness properties of the pressurizer; determine that the pressurizer remains acceptable for continued opera-r,

/

tion 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 reduce the pressurizer pressure to less than 500 psig within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIhtNENTS 4.4.8.2 The pressurizer temperatures shall be determined to be within the limits:

a.

At least once per 30 minutes during system heatup or cooldown, and l

b.

By verifying the Main Coolant System and pressurizer water temperature differential to be within the limit at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during steady state operation.

YANKEE-ROWE 3/4 4-25 l

l Amendment No. Aff 49 l

l ll ---

MAIN COOLANT SYSTEM STRUCTURAL INTEGRITY CLASS I COMP 0NENTS

~

LIMITING CON 0! TION FOR OPERATION y

,j-7,'..

E 3.4.9 The structural integrity of Main Coolant System components (except steam generator tubes) identified in Table a.4-3 as Class 1 components shall be maintained at a level consistent with the acceptance criteria

~

in Specific.tions 4.4.9.1, 4.4.9.2, 4.4.9.3 and 4.4.9.4.

A_PptICABILITY: ' NODES 1, 2, 3 and 4 ACTIO_M_ :

With the structural f ategrity of any of the above components not con-forming to the above requirments, restore the structural integrity of the affected component to within its limit or isolate the affected component prior to increasing the Main Coolant System temperature more than 50*F above the minimum temperature required by NOT considerations.

l7 The provisions of Specification 3.0.4 are not applicable.

s SURVEILLANCE REQUIREMENTS s 4.4. 9.1 The following inspection program shall be performed during

/

shutdown:

a.'

Inservice' Inspections The structural integrity of the Class f

I components s. hall be demonstrated by verifying their accept-ability whe6 inspected per the applicable requirements of V

Section XI of the ASME Boiler and Pressure Vessel Code,1970 Edition, and Addenda through Winter 1970, as outlined by the inspection program shown in Table 4.4-3.

~

For all Class I ptptng the ultrasonic calibration shall be per:

~

1.

Article III-200 of Appendix III - ASME Sec XI-Summer 1976 Addenda except that III-2410 shall be deleted III-2430 shall be used except 50% Reference level recording shall be performed. Ten percent overlap shall be retained.

~

Article III-3000 shall be used entirely.

2.

3.

Article III-4000 shall be used entirely.

4.-

Supplement 7 shall be used for austenttic welds.

YANKEE-ROWE 3/4 4-26 r

i EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (continued) b.

At least once per 31 days by:

1.

Verifying that the following valves are in the indicated positions with power to the valve operators removed by coening at least two breakers in series:

Valve Number Valve Function Valve Position

a. CH-MOV-522

a. Charging Header /LPSI

a. Closed Isolation

b. CH-MOV-523 b.&c. Charging Header / Loop

b. Open 4 Hot Leg Injection

c. CH-M0V-524 Long-Term

c. Opcn Recirculation 2.

Cycling each 'of the following valves, except PU-V-651, through at least one complete cycle of full travel and verifying that the valves are in the indicated normai positions:

Valve No.

Valve Position PU-MOV-541 Open PU-MOV-542 Open PU-MOV-543 Closed PU-MOV-544 Closed PU-MOV-b45 Open PU-M0V-546 Open PU-M0V-547 Closed PO-MOV-548 Closed PU-V-651 Open 3.

Verifying that the following valves are in their normally opened positions with power to the valve operators removed by removal of the circuit breaker from the motor control center:

Valve riumber Valve Function

a. SI-MOV-4

a. LPSI Pump Cross Over to HPSI Pump

b. SI-MOV-22

b. 51 Header Isolation to Cold Leg

c. SI-MOV-23

c. SI Header Isolation to Cold Leg

d. SI-MOV-24

d. SI Header Isolation to cold Leg l

e. SI-MOV-25

e. SI Header ! solation to Cold Leg i

f. SI-MOV-46

f. HPSI Flow Control

g. SI-MOV-49

g. liPSI Recirculation Test Line YANKEE-ROWE 3/4 5-5 i

^

EMERFENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (continued) 4.

Verifying that power to the valve operators is removed by disconnecting the power cables as they leave the motor starters:

Valve Number Valve Function

a. CS-MOV-533

a. LPSI Pump Header Isolation

b. CS-MOV-535

b. LPSI Pump Header Isolation

c. CS-MOV-536

c. SI Header Isolation to Cold Leg

d. CS-MOV-537

d. SI Header Isolation to Cold Leg

e. CS-MOV-538

e. SI Header Isolation to Cold Leg

f. CS-MOV-539

f. SI Header Isolation to Cold Leg

g. MC-MOV-301

g. MCS Loop _ Isolation

h. MC-MOV-302*

h. MCS Loop -Isolation l

i. MC-MOV-309

1. MCS Loop Isolation

j. MC-MOV-310*

j. MCS Loop Isolation

k. MC-MOV-318*

k. MCS Loop Isolation

1. MC-MOV-319

1. MCS Loop Isolation

m. MC-MOV-325

m. MCS Loop Isolation

n. MC-MOV-326*

n. MCS Loop Isolation l

S.

Verifying that the following valve is in its normally closed position with power to the valve operator removed 4

by disconnecting the power cables as t.iey leave the motor i

starter:

Valve Number Valve Function

a. CS-MOV,-532

a. LPSI Recirculation Line Note: This valve may be opened for < 30 minutes once per week for safety injection tank mixing or low pressure safety injection pump testing after restoring power to the valve operator.

Insure that power to the valve operator is properly removed after closing the valve.

6.

Verifying that each ECCS safety injection subsystem is aligned to receive electrical power from an OPERABLE emergency bus.

In MODE 2, 3*,

4* or 5*, power cables may be connected to the MCS loop isolation valves when required to close the valves for main coolant

'3 pump (s) starting. After the pump (s) has been started, the valve (s) shall been reopened and power cables disconnected.

YANKEE-ROWE 3/45-6 Amendment No. 49

.ev*-

p--w-y 9s--am.i->v.gy-w,v-i.m,a.

,-ii-..ai.e

-e--y 9 >ggwga 1.e, y

ggn

-w

-+v,p,v---r ny-vg--m-ww--

--wva-m----

w.

u EMERGENCY CORE COOLING SYS1 3S ECCS SUBSYSTEMS LIMITING CONDITION FOR OPERATION 3.5.3 The ECCS subsystems shall k OPERABLE with:

a.

As a minimum, one OPERABLE ECCS safety injection subsystem comprised of the following:

1.

One OPERABLE high pressure safety injection pump, 2.

One OPERABLE low pressure safety injection pump, and 3..

An OPERABLE flow path capable of taking suo..on from the safety injection tank.

b.

An OPERABLE ECCS recirculation subsystem with at least one OPERABLE purification pump and an OPERABLE flow path capable of taking suction from the containment sump and recirculating to the safety injection header.

c.

An OPERABLE ECC long term hot leg injection subsystem with at least one OPERABLE fixed speed charging pump and an OPERA 8LE flow path capable of taking suction from the ECCS ecircu-lation subsystem and discharging to the Main Coolant System

4 hot leg.

~

I, APPLICABILITY: MODE 4*.

ACTION:

' With the ECC'S afety injection subsystem, the recirculation a.

subsystem, or the long term hot leg injection subsystem in-l operable, restore all subsystems to OPERABLE status within 1 l

hour or be in COLD SHUTDOWN within the next 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .

b.

In the event the ECCS is actuated and injects water into the Main Coolant Systcm, a Special Report shall be prepared and submitted to the Comission pursuant to Specification 6.9.6 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date.

Main coolant pressure < 1000 psig but > 300 psig.

YANKEE-ROWE 3/4 5-9 Amendment No. 49 l

t EMERGENCY t.0RE COOLING SYSTEMS LCCS SUBSYSTEMS SURVEILLANCE REQUIREMENTS 4.5.3 The ECCS subsystems shall be demonstrated OPERABLE per the ap-plicable Surveillance Requirement; of 4.5.2.

L a

e YANKEE-ROWE 3/4 6-10 S

EMERGENCY CORE COOLING SYSTEMS SAFETY INJECTION TANK LIMITING CONDITION FOR OPERATION 3.5.4 The safety injection tank (SIT) shall be OPERABLE with:

a.

A minimum contained borated water volume of 117,000 gallons, equivalent to a level of,25.5 feet.

3 b.

A minimum boron concentration of 2200 ppm, and c.

A minimum water temperature of 4G'F.

APPLICABILITY: MODES 1, 2, 3, 4 and 5*

ACTION:

With the safety injection tank inoperable, restore the tank to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN with Main Coolant pressure < 1000 psig within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.5.4 The SIT shall be demonstrated OPERABLE:

a.

At least once per 7 days by:

1.

Verifying the contained barated water volume in the tank, and 2.

Verifying the boron concentration of the water.

b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the SIT temperature when the outside air temperature is < 35'F.

Main coolant pressure 3,1000 psig, YANKEE-ROWE 3/4 5-11 s

Amendment No. 49

3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained.

APPLICABILITY: MODES 1, 2, 3, 4 and 5*.

ACTION:

Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within one hour or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN with Main Coolant pressure < 300 psig within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:

a.

At least once per 31 days by verifying that:

1.

All penetrations not capable of being closed by OPERABLE containment automatic isolation valves ;nd required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured in their positions, except as provided in Table 3.6-1 of Specification 3.6.2, and 2.

All equ'ipment hatches are closed and sealed.

b.

By verifying that the containment air lock is OPERABLE per Specification 3.6.1.3.

By verifying that the containment continuous leak monitoring c.

system is OPERABLE per Specification 3.6.1.7.

d.

By calculating and plotting the containment air mass as deter-mined by the containment continuous leak monitoring system at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Main coolant pressure > 300 PSIG YANKEE-ROWE 3/4 6-1 Amendment No. 49

CONTAINMENT SYSTEMS s

I

' CONTAINMENT LEAKAGE LIMITING CONDITION FOR OPERATION 3.6.1.2 Containment leakage rates shall be limited to:

a.

An overall integrated leakage rate of:

1 a, 0.20 percent by weight of the containment air per 1.

L 24 hours at P,, 31.6 psig, or 1 g, 0.1123 percent by weight of the containment air per L

2.

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at a reduced pressure of P.16 psig.

t b.

A combined leakage rate of < 0.60 L for all penetrations and valves subject to Type B and C tests as identified in Table 3.6-1, when pressurized to P,.

I APPLICABILITY: MODES 1, 2, 3, 4 and 5*

ACTION:

With either (a) the measured overall integrated containment leakage rate or 0.70 L., as applicable, or (b) with the measured exceeding 0.70 L, rate for all penetrations and valves subject to Types B combined leakage and C tests exceeding 0.60 L,, restore the overall integrated leakage all penetration $ or <0.70 L as applicable, and the combined rate for rate :o 10.70 L t

increasingtheMainCoolantSystemtemperatureabove200*f.priorto subTect to Type B and C tests to 10.60 L SURVEILLANCE REQUIREMENTS 4.6.1.2 The containment leakage rates shall be demonstrated at the following test schedule and shall be detemined in conformance with the criteria specified in Appendix J of 10 CFR 50 using the methods and provisions of ANSI N45.4-(1972):

Three Type A tests (Overall Integrated Containment Leakage a.

Rate) shall be conducted at 40 + 10 month intervals during shutdown at either P, 31.6 psig, or at P, 16 psig, during each 10-year service period. The third thst of each set shall a

be conducted during the shutdown for the 10-year plant in-service inspection.

Main Coolant pressure > 300 psig.

YANKEE-ROWE 3/4 6-2 Amendment No.49 h

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

If any periodic Type A test tails to meet either.75 L or b.

.75 L, the test schedule for subsequent Type A tests $ hall be reviewed and approved by the Commission.

If two consecutive Type A tests fail to meet either 0.70 L or 0.70 L, a Type A test shall be perfonned at least every i8 months u Itil two consecutive Type A tests meet either 0.70 L or 0.70 L at t

which time the above test schedule may be r$sumed.

c.

The accuracy of each Type A test shall be verified by a supplemental test which requires the metered mass of gas injected into the containment or bled from the containment for the supplemental test to be equivalent to between 60 and 100% of the allowable 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months mass loss. The acceptability is demonstrated if the mass change, as measured by the Type A instrumentation, agrees with the mass change as metered by the flow meter to within 25% of the allowable 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months mass loss.

Type B and C tests shall be conducted with gas at P, 31.6 psig, d.

at intervals no greater than 24 months except for t$sts involving the air lock, the equipment hatch, the emergency hatch, the containment leg expansion Joints, and the fuel chute expansion joint.

e.

The air lack shall be tested and demonstrated OPERABLE per Specification 4.6.1.3.

f.

The equipment and emergency hatch seals and seating surfaces shall be inspected before each hatch closure.

g.

All test lea'kage rates shall be calculated using observed data converted to absolute values. Error analyses shall be performed to provide a maximum expected error.

l 1

l l

YANKEE-ROWE 3/4 6-3 l

CONTAINMENT SYSTEMS CONTAINMENT AIR LOCK LIMITING CONDITION FOR OPERATION 3.6.1.3 The containment air lock shall be OPERABLE with:

a.

Both doors closed except when the air lock is being used for normal transit entry and exit through the containment, then at least one air lock door shall be closed, and b.

An overall air lock leakage rate of < 0.05 L at P,, 31.6 a

psig.

APPLICABILITY: MODES 1, 2, 3, 4 and 5*.

ACTION:

With the air lock inoperable, restore the air lock to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.6.1.3 The containment air lack shall be demonstrated OPERABLE:

By verifying that the containment continuous leak monitoring a.

system is OPERABLE per Specification 3.6.1.7.

b.

By calculating and plotting the containment atmosphere mass as determined by the containment continuous leak monitoring system at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

By inspecting the door seals and seating surface after each c.

opening, except when the air lack is being used for multiple entries, then at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , d.

At least once per 6 months by conducting an overall air lock leak:ge test at P 31.6 psig, and by verifying that the overallairlockie,akagerateiswithinitslimit,and

... least once per 6 months by verifying that only one joor in l

e.

the air lock can be opened at a time.

l

Main Coolar.t pressure > 300 psig.

YANKEE-ROWE 3/46-4 Amendment No. 49 e-

+

I

\\

i CONTAINMENT SYSTEMS CONTAINMENT V5SSEL_ STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION t

3.6.1.6 The structural integrity of the containment vessel shall be maintained at a level consistent with the acceptance criteria in Specification 4.6.1.6.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With the structural integrity of the containment vessel not conforming to the above requirements, restore the structural integrity to within the limits prior to increasing the Main Coolant System temperature above 200kF.

SURVEILLANCE REQUTREMENTS 4.6.1.6 The structural integrity of the containment vessel shall be determined during the shutdown for each Type A containment leakage rate test (reference Specification 4.6.1.2) by a visual inspection of the accessible interior and exterior surfaces of the vessel and verifying no apparent changes in appearance of the surfaces or other abnormal degradation.

An initial report of any. abnormal degradation of the containment vessel detected during the above required inspections shall be made within 10 days after detection and the detailed report shall be submitted to the Commission pursuant to Specification 6.9.1 within 90 days after completion of the surveillance requicements of this specification.

YANKEE-ROWE 3/4 6-7

CONTAINMENT SYSTEMS CONTINUOUS LEAK MONITORING SYSTEM LIMITING CONDITION FOR OPERATION 3.6.1.7 The continuous leak monitoring system shall be OPERABLE within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months following establishment of CONTAINMENT INTEGRITY with:

a.

Containment internal pressure > 0.75 psig, b.

At least eight containment temperature detectors, c.

At least one containment pressure detector, d.

Two relative humidity detectors.

APPLICABILITY: MODES 1, 2, 3, 4 and 5*

l ACTION:

With the continuous leak monitoring system inoperable, restore the system to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after closing any containment air lock door, whichever is sooner, or be in at least P0T STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN with Main Coolant pressure < 300 psig within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.6.1.7 The continuous leak monitoring system shall be demonstrated OPERABLE by:

i Verifying containment internal pressure to be > 0.75 psig at a.

l least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

1 b.

Calibrating the temperature detectors, the pressure detector (s) l and the relative humidity detectors at least once per 18 months.

Main coolant pressure > 300 psig.

l l

YANKEE-ROWE 3/4 6-8 Amendment No. 49

CONTAINMENT SYSTEMS _

3/4.6.2 CONTAINMENT ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6.2 The containment isolation valves specified in Table 3.6-1 shall be OPERABLE with isolation times as shown in Table 3.6-1.

APPLICABILITY: MODES 1, 2, 3, 4 and 5*.

ACTION:

With one or more of the isolation valve (s) specified in Table 3.6-1 inoperable either:

Restore the inoperable valve (s) to OPERABLE status eithin 4 a.

hours, or b.

Isolate each affected penetration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least one deactivated automatic valve secured in the isolation position, or Isolate each affected penetration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at c.

least one closed manual valve or blind flange; or d.

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 .

SURVEILLANCE REQUIREMENTS 4.6.2.1 The isolation valves specified in Table 3.6-1 shall be demon-strated OPERABLE:

At least once per 92 days by cycling each OPERABLE power a.

operated or automatic valve testable during plant operation through at least one complete cycle of full travel.

Imediately prior to returning the valve to service after b.

maintenance, repair or replacement work is performed on the valve or its associated actuator, control or power circuit by perfcrmance of the cycling test, above, and verification of isolation time.

Main coolant system pressure > 300 psig.

YANKEE-ROWE 3/4 6-9 Amendment No. 49

CONTAINMENT SYSTEMS s

SURVEILLANCE REQUIREMENTS (Continued) 4.6.2.2 Each isolation valve specified in Table 3.6-1 shall be demon-strated OPERABLE during the COLD SHUTDOWN or REFUELING MODE at least once per 18 months by:

Verifying that on a containment isolation tes+ signal, each a.

automatic isolation valve actuates to its iso. tion position, b.

Cycling each power operated or automatic valve through at

-j least one complete cycle of full travel and measuring the isolation time, and Cycling each manual valve not locked, sealed or otherwise l

c.

secured in the closed position through at least one complete cycle of full travel.

t I

YANKEE-ROWE 3/4 6-10 e-Amendment No. 43 3

.~

TABLE 3.6-1 (Continued)

CONTAINMENT ISOLATION VALVES 5jii TESTABLE DURING E

VALVE NUMBER FUNCTION PLANT OPERATION ISOLATION TIME (Yes or No)

(Seconds)

E B.

CHECK VALVES (Continued)

SW-V-820*

Service Water to Containment Cooler #1 NA NA 5W-V-821*

Service Water to Containment Cooler #2 NA NA SW-V-822*

Service Water to Containment

- Cooler #3 NA NA SW-V-823*

Service Water to Ccntainment Cooler #4 NA NA HC-V-1199*

Steam F epply to Containment Heaters NA NA 2

NA C.

Manual Valves h

SC-MOV-551+553*

toown Cooling - In No NA SC-MOV-552+554*

Shutdown Cooling - Out No NA CH-MOV-522*

MC Feed to Loop Fill Header NA NA CS-V-601 Shield Tank Cavity Fill NA NA CA-V-746*

Containment Air Charge NA NA HV-V-5*

Containment H2 Vent System NA NA HV-V-6*

Containment H2 Vent System NA NA CA-V-688 Containment H2 Vent System Air Supply NA NA CS-M0V-500 Fuel Chute Lock Valve No NA

$Not subject to Type C tests

5 TABLE 3.6-1 (Continued).

CONTAINMINT ISOLATION VALVES j

n lE TESTABLE DURING la E

VALVE NUMBER FUNCTION PLANT OPERATION ISOLATION TIME (Yes or No)

Seconds C.

Manual Valves (Cont'd)

CS-CV-215 Fuel Chute Equalizing NA NA CS-CV-216 Fuel Chute Dewatering NA NA Pump Discharge VD-V-752*

Neutron Shield Tank-Outer Test NA NA VD-V-754*

Neutron Shield Tank-Inner Test NA NA w

BF-V-4-1 Air Purge Inlet NA NA NA NA 3

BF-V-4-2 Air Purge Outlet NA NA HC-V-602 Air Purge Bypass PU-H0V-543 ECCS Recirculation Yes 50 PU-MOV-544 ECCS Recirculation Yes 50 g

BF-CV-1000*

SG#1 Feedwater Regulator No 30 BF-CV-1100*

SG#2 Feedwater Regulator No 30 g

BF-CV-1200*

SG#3 Feedwater Regulator No 30 m

BF-CV-1300*

SG#4 feedwater Regulator No 30 g

PR-V-623 Main Coolant Heise Pressure Gauge NA NA l

8

Not subject to Type C tests W

-W

PLANT SYSTEMS TURBINE GENERATOR THROTTLE AND CONTROL VALVES LINITING CONDITION FOR OPERATION 3.7.1.5 Each turbine generator throttle and control valve shall be OPERABLE.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

MODES 1 - With one turbine generator throttle or control valve inoperable, POWER OPERATION may continue provided the inoperable valve is either restored to OPERABLE status or closed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; Otherwise, be in at least HOT SHUTDOWN within the next 12 l

hours.

~

MODES 2 - With one turbine generator throttle or control valve inoperable, and 3 subsequent operation in MODES 1, 2 or 3 may proceed provided the inoperable valve is maintained closed; otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.1.5 Each turbine generator throttle and control valve that is open shall be-demonstrated OPERABLE by:

Cycling each valve through at least one complete cycle of full a.

travel at least once per month, and 1

b.

Verifying full closure within 2 seconds on any closure actuation signal whenever shutdown longer than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , if not performed in the previous 92 days.

i YANKEE-ROWE 3/4 7-9 Amendment No. 49 i

e i

t.

PLANT SYSTEMS SECONDARY WATER CHEMISTRY LIMITING CONDITION FOR OPERATION 3.7.1.6 The secondary water chemistry shall be maintained within the limit of Table 3./-3.

APPQCABILITY: fiODES 1, 2 and 3.

ACTION:

a.

With the secondary water chemistry parameter outside of its limit restore the parameter to within its limit within 3 days; otherwise be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in CC2.0 SHUTDOWN within the following 96 nours.

b.

The provisians of Specification 3.0.4 are not applicable for up to 72 haue s provided:

1.

The chloride concentration is < 2.0 ppm, 2.

the THERMAL POWER is-< 30% of RATED IHERMAL POWER, and 3.

Corrective measures have been implemented to restore chloride concentrations to < 0.5 ppm.

SURVEILLANCE REQUIREMENTS 4./.l.6 The secondary water chemistry shall be determined to oe within the limit by analysis of the parameter at the frequency specified in Table 4.7-3.

YANKEE-ROWE 3/4 7-10

PLANT SYSTEMS TURBINE GENERATOR THROTTLE AND CONTROL VALVES LIMITING CONDITION FOR OPERATION 3.7.1.5 Each turbine generator throttle and control valve shall be OPERABLE.

APPLICA6ILITY: MODES 1, 2 and 3.

ACTION:

MODES 1 - With one turbine generator throttle or control valve inoperable, POWER OPERATION may conthiue provided the inoperable valve is either restored to OPERABLE status or closed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; Otherwise, be in at least HOT SHUTDOWN within the next 12 l

hours.

MODES 2 - With one turbine generator throttle or control valve inoperable, and 3 subsequent operation in MODES 1, 2 or 3 may proceed provided the inoperable valve is maintained closed; otherwise, be in at

.least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.1.5 Each turbine generator throttle and control valve that is open shall be demonstra.3d OPERABLE by:

a.

Cycling each valve through at least one complete cycle of full travel at least once per month, and b.

Verifying full closure within 2 seconds on any closure actuation signal whenever shutdown longer than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , if not perforned in the previous 92 days.

4 YANKEE-ROWE 3/4 7-9 Amendment No. 49

PLANT SYSTEMS SECONDARY WATER CHEMISTRY LIMITING CONDITION FOR OPERATION 3.7.1.6 The secondary water chemistry shall be maintained within the limit of Table 3./-3.

APPLICABILITY: it0 DES 1, 2 and 3.

ACTION:

a.

With the secondary water chemistry parameter outside of its limit, restore the parameter to within its limit within 3 days; otherwise be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 96 nours.

b.

The provisions of Specification 3.0.4 are not applicable for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months provided:

1.

The chloride concentration is < 2.0 ppm, 2.

the THERMAL POWER is < 30% of RATED IHERMAL POWER, and 3.

Corrective measures have been implemented to restore chloride concentrations to < 0.5 ppm.

SURVEILLANCE REQUIREMENTS 4./.l.6 The secondary w ter chemistry shall be determined to be within the limit by analysis of the parameter at the frequency specified in Table 4.7-3.

YANKEE-ROW'i 3/4 7-10

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) c.

At least once per 18 months during shutdown, by verifying that the standby service water pump starts automatically on a dis-charge header pressure > 40 psig.

l f

YANKEE-ROWE 3/4 7-17 Amendment No. 49

s PLANT SYSTEMS 3/4.7.5 CONTROL ROOM VENTILATION SYSTEM EMERGENCY SHUTDOWN LIMITING CONDITION POR OPERATION 3.7.5 The control room ventilation system emergency shutdown shall be OPERABLE.

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

ACTION:

With the control room ventilation system emergency shutdown inoperable, restore the system to OPERA 3LE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least HOT 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 .

SURVEILLANCE REQUIREMENTS 4./.5 The control room ventilation system emergency shutdown shall be demonstrated CPERABLE at least once per 31 days by verifying that the control room ventilation system can be shutdown manually f rom the control room.

YANKEE-ROWE 3/4 7-18

PLANT SYSTEMS 3/4.7.7 WASTE EFFLUENTS RADIOACTIVE SOLIO WASTE LIMITING CONDITION FOR OPERATION 3.7.7.1 Radioactive solid waste shall not be disposed of at the site.

l APPLICABILITY: At all times ACTIGN:

With any radioactive solid waste disposed of at the site, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.6 within 90 days describing the circumstances of the disposal cnd outlining plans for removal of the waste. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.7.1 Not applicable.

l YANKEE-ROWE 3/4 7-21 Amendment No. 49

PLANT SYSTEMS RADI0 ACTIVE LIQUID WASTE LIMITING CONDITION FOR OPERATION 3.7.7.2 Radioactive liquid waste shall be discharged only when the activity of the waste, together with the activity being released from steam generator blowdown, is less than the maximum permissible concen-tration established in 10 CFR Part 20.

APPLICABILITY: At all times ACTION:

With discharge of radioactive liquid waste in excess of the limits, immediately suspend the discharge. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS l

4.7.7.2.1 Radioactive liquid waste shall be determined to De within the above limits by radioactivity analysis prior to discharge, t

i 4.7.7.2.2 Steam generator blowdown radioactivity shall be analyzed at least once every 7 days whenever any steam generator contains water.

YANKEE-ROWE 3/4 7-22

PLANT SYSTEMS RADI0 ACTIVE GASEOUS WASTE LIMITING CONDITION FOR OPERATION 3./.7.3 Ccncentration of radioactive gaseous wastes discharged, as determined at the point of discharge from the primary vent stack and averaged over a period not exceeding one year, shall not exceed 1000 times the limits specified in 10 CFR Part 20, Appendix B. Table II, except that, for isotopes of Iodines and particulates with half-lives

> 8 days, the MPC values of Table II shall be reduced by a factor of 1/700.

APPLICABILITY: At all times ACTION:

With discharge of radioactive gaseous waste in excess of limits, im-mediately suspend the discharge.

The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.7.3 Radioactive gas'eous waste discharge concentrations shall be determined to be within the limits, by continuously monitoring the primary vent stack effluent and analyzing for tritium, particulates, iodines, and tission, and activation gases at least once per 31 days.

YANKEE-ROWE 3/4 7-23

s PLANT SYSTEMS 3/4.7.8 ENVIRONMENTAL MONITORING LIMITING CONDITION FOR OPERATION 3.7.8 The environmental monitoring program shall be performed in accordance with Table 4.7-4.

APPLICABILITY: At all times ACTION:

With the sampling and analysis program specified in Table 3.7-4 not satisfied, a special report shall be prepared and submitted to the Commission pursuant to Specification 6.9.6 within 90 days describing the l circumstances of the violation and outlining plans to prevent re-occur-rence of the violation. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.8 The environmenthl' monitoring samples shall be collected and analyzed in accordance with the requirements of Table 4.7-4.

l t

YANKEE-ROWE 3/4 7-24 Amendment No. 49 l

[

1 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 with:

1.

Day fuel tank containing a minimum volume of 210 gallons of ruel, equivalent to a 3/4 full tank, and 2.

A fuel storage system containing a minimum volume of 4000 gallons of fuel, equivalent to a tank level of 2'2".

APPLICABILITY: MODES 5 and 6.

ACTION:

With less than the above minimum required A.C. electrical power sources OPERABLE, suspend all cperations involving CORE ALTERATIONS or positive reactivity changes until the minimum required A.C. electrical power sources are restored to OPERABLE status.

SURVEILLANCE REQUIREMENIS 4.8.1.2 The above required A.'C. electrical power sources shall be demonstrated OPERABLE by the performance of each of the Surveillance Requirements of 4.8.1.1.1 and 4.8.1.1.2 except for requirement 4.8.1.1.2a.4.

rANKEE-ROWE 3/4 d-6

ELECTRICAL POWER SYSTEMS s.

3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS A.C. DISTRIBUTION - OPERATING LIMITING CONDITION FOR OPERATION 3.8.2.1 The folicwing A.C. electrical busses shall be OPERABLE and -

energized from sources of power other than the diesel generators with tie breakers open between redundant busses:

a.

Redundant busses:

1.

2400 volt bus #1 2.

2400 volt bus =2 3.

2400 volt cus 43 4.

480 volt bus d4-l 5.

480 volt bus #f-2 6.

480 volt bus #6-3 7.

480 volt Emergency Bus #1 8.

480 volt Emergency Bus #2 9.

480 volt Emergency Bus #3 l

10.

480 volt MCC al bus 2.

b.

Non-Redundant Susses:

1.

480 volt Emergency MCC #1 2.

480 volt Emergency MCC #2 3.

120 volt Vital Bus.

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

With less than the above canplement of redundant A.C. busses OPERABLE, a.

restore the inoperable bus to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be

One tie breaker may be closed < 30 MWe.

YANKEE-ROWE 3/4 8-6 Amendment No. 49

ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION (Continued) 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 .

b.

With any non-redundant bus inoperable, be in at least HOT STANDBY within one hour and in COLO SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.8.2.1.1 Each specified A.C. bus and emergency MCC shall be determined OPERABLE and energized from A.C. sources other than the diesel gen-erators with tie breakers open between redundant busses at least once per 7 days by verifying correct breaker alignment and indicated power availability.

4.8.2.1.2 Onergency MCC #1 and #2 busses shall be determined OPERA 8LE by verifying:

a.

At least once per 31 days that the alternate power supply is disconnected by racking out and locking out the breakers, and b.

At least once per 18 months that the interlocks preventing the normal and alternate breakers from simultaneously being in the closed' position are OPERABLE.

4.8.2.1.3 The 120 volt vital bus shall be determined OPERABLE at least once per 18 months by manually transferring vital bus power from the normal source to each of the following alternate power sources:

a.

480 volt Emergency MCC #1.

b.

480 volt MCC #1 bus 2.

YANKEE-ROWE 3/4 8-7 Amendment No. 49 s

7

-w ELECTRICAL POWER SYSTEMS A.C. DISTRIBUTION - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.2.2 As a minimum, the following A.C. electrical busses shall be OPERABLE and energized from sources of power other than a diesel gen-erator but aligned to an OPERABLE diesel generator.

a.

1 - 2400 volt bus #2 or #3 b.

2 - 480 volt buses #4-1 and #5-2 1 - 480 volt Emergency Buses #1, 2 or 3 c.

d.

2 - 480 volt buses Emergency MCC *1 and Emergency MCC #2 e.

1 - 120 volt Vital Bus APPLICA8ILITY: MODES 5 and 6.

ACTION:

With less than the above complement of A.C. busses OPERABLE and ener-gized, establish CONTAINMENT INTEGRITY within 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months .

SURVEILLANCE REQUIREMENTS 4.8.2.2 The specified A.C. busses shall be determined OPERABLE and energized from A.C. sources other than the diesel generators at least once per 7 days by verifying correct breaker 111gnment and indicated power availability.

YANKEE-ROWE 3/4 8-0

l SPECIAL TEST EXCEPTIONS PRESSURE / TEMPERATURE LIMITATION

, REACTOR CRITICALITY LIMITING CONDITION FOR OPERATION 3.10.'

The minimum temperature and pressure conditions for reactor critical-ity of Specification 3.4.3.1 may be suspended diaring low temperature PHYSICS TESTS provided:

a.

The THERMAL POWER does not exceed 2 percent of RATED THERMAL

POWER, b.

The reactor low setpoint trips on the three OPEI ABLE Power Range Nuciear Channels are set at <25%.of RATED THERMAL POWER, and The Main Coolant System temperature and pressure are maintained c.

3,250*F and 3,300 psig, respectively.

APPLICABILITY: MODE 2.

ACTION:

With the THERMAL POWER > 2 percent of RATED THERMAL POWER, a.

immediately open the reactor trip breakers.

b.

With the ' lain Coolant System temperature and pressure < 250*F or < S00 psig, immediately open the reactor trip breakers and

restore the temperature-pressure to within its limit within 30 minutes; perform the analysis required by Specification 3.4.8.1 prior to the next reactor criticality.

SURVEILLANCE REQUIREMENTS 4.10.3.1 The Main Coolant System emperature and pressure shall be verified to be 3,250*F and 300 psig at least once per hour.

l 4.10.3.2 The THERMAL POWER shall be determined to be < 2% of RATED THERMAL POWER at least once per hour.

4.10.3.3 Each Pow:r Range Nt clear Channel shall be subjected to e CHANNEL FUNCTIONAL TEST within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months prior to initiating low tempera-ture PHYSICS TESTS.

3/4 10 3 Anendment No. /49

. YANKEE-ROWE

SPECIAL TEST EXCEPTIONS PHYSICS TESTS LIMITING CONDITION FOR OPERATION 3.10.4 The limitations of Specification 3.1.1.4, 3.1.3.1, 3.1.3.4, and 3.1.3.5, may be suspended during the performance of PHYSICS TESTS provided:

a.

The THERMAL POWER does not exceed 2% of RATED THERMAL POWER, and o.

The reactor low setpoint trips on the three OPERABLE Power Range Nuclear Ch;nnels are set at < 25% of RATED THERMAL POWER.

APPLICABILITY: MODE 2.

ACTION:

With the THERMAL POWER > 2% of RATED THERMAL POWER, immediately open the reactbi' trip breakers.

SURVEILLAMCE REQUIREMENTS 4.10.4.1 The THERMAL POWER shall be determined to be < 2% of RATED THERMAL POWER at least once per hour during PHYSICS TETTS.

~

4.10.4.2 Each Power Range Nuclear Channel shall be subjected to a l

CHANNEL FUNCTIONAL TEST within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months prior to initiating PHYSICS TESTS.

YANKEE-R0WE 3/4 10-4 Amendment No. 47

3/4.4 MAIN C00UlNT SYSTEM BASES 3/4.4.1 MAIN COOLANT LOOPS The plant is designed to cperate with all main coolant loops in operation, and maintain DNBR above 1.30 during all nonnal operations and anticipated transients. 'dith one main coolant loop not in operation, THERMAL POWER is restricted to < 75 percent of RATED THERMAL POWER.

With four loops operating, a loss of flow or low SG water level in two loops will cause a reactor trip. A loss of flow or low SG water level in one loop will cause a reactor trip with three loops operating.

Adequate main coolant loops are required to provide sufficient heat removal capability for removing core decay heat. Single failure con-siderations require placing the Shutdown Cooling System into operation if the required main coolant loops are not OPERABLE. A steam generator is capable of removing core decay heat by natural or forced circulation provided the conditions specified in 4.4.1.1.2 are met.

The requirement to maintain the baron concentration of an isolated loop greater than or equal to the boron concentration of tne operating loops ensures that no reactivity additier. to the core could occur during startup of an isolated loop. Verification of the boron concentration in an isolated loop prior to opening the stop valves provides a reassurance of the adequacy of the boron concentration in the isolated loop. Startup of an isolated loop will inject cool water from the loop into the core.

The reactivity transient resulting from this cool water injection is minimized by delaying isolated loop startup until its temperature is within 30*F of the operating loops. Making the reactor subcritical l

prior to loop startup prevents any power spika which could result from this cool water induced reactivity transient.

3/4.4.2 and 3/4.4.3 SAFETY VALVES The pressurizer code safety valves operate to p*avant the Main Coolant System from being pressurized above its Safety Limi' of 2735 psig. Each safety valve is designed to relieve 92.000 lbs per hour of saturated steam at the valve set point. The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during sNtdown.

In the event that no safety valves are OPERABLE, an operating Shutdown Cooling System connected to the Main l

Coolant System provides overpressure relief capability and will prevent Main Coolant System overpressurization during shutdown.

(

l

' YANKEE-ROWE B 3/4 4-1 Amendment No. 49 i

-~

MAIN COOLANT SYSTEM BASES Uuring operation, all pressurizer code safety valves must be OPERABLE to prevent the Main Coolant System trom being pressurized above its safety limit of 2735 psig. The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from a complete loss of load assuming no reactor trip until the first Reactor Protective System trip set point is reached (i.e., no credit is taken for a direct reactor trip on the loss of load) and also assuming no operation of the power operated relief valves or steam dump 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.

3/4.4.4 PRESSURIZER A steam bubble in the pressurizer ensures that the Hain Coolant System is not a hydraulically solid system and is capable of accom-modating pressure surges during operation. The steam bubble also protects the pressurizer code, safety valves and power operated relief valve against water relief. The power operated relief valve and steam bubble function to relieve Hain Coolant System pressure during all design transients up to and including the design step load decrease with steam dump. Operation of the power operated relief valve minimizes the undesirable opening of the spring-loaded pressurizer code safety valves.

YANKEE-ROWE B 3/4 4-2

MAIN COOLANT SYSTEM _

BASES Ouring heatup, the thermal gradients in the reactor vessel wall produce thermai stresses which vary from compressive at the inner wall to tensile at the outer wall. These thennal induced compressive stresses tend to alleviate the tensile stresses induced by the internal pressure.

Therefore, a pressure-temperature curve based on tteady state conditions (i.e., no thennal stresses) represents a lower bound of all similar curves for finite neatup rates when tne inner wall of the vessel is treated as the governing location.

The heatup analysis also covers the detennination of pressure-temperature liuitations for the case in which the outer wall of the vessel becomes the controlling location. The thermal gradients established during heatup produce tensile stresses at the outer wall of the vessel. These stresses are additive to the pressure induced tensile stresses which are already present. The thermal induced stresses at the outer wall of the vessel are tensile and are dependent on both the rate of heatup and the time along the neatup camp; therefore, a lower bound cur,e similar to that described for the heatup of the inner wall cannot be defined. Consequently, for the cases in which the outer wall of the vessel becomes the stress controll'ng location, each heatup rate of interest must be analyzed on an individual basis.

The heatup limit curves, Figures 3.4-2, and 3.4-4, are composite curves which were prepared by detennining the most conservative case, with either the inside or outside wall controlling, for any heatup rate up to 100*F per hour. The cooldown limit curves, tigures, 3.4-3, and 3.4-5, are composite curves which were prepared based upon the same type analysis. The heatup and cooldown curves were prepared based upon a 10'F.

These limitations are derived by using beginning of life RTthe rules contained NT_e+ction III of the ASME Code including Ap S

G. Protection Against Nonductile Failure, and the rules contained in 10CFR50, Appendix u, tracture Toughness Requirements.

Reactor operation and resultant fast neutron (E>l Mev) irradiation controlling vessel material E. The weld metal is assumed to be the will cause an increase in RT ughout the remainder of vessel life.

The shift in RT ca.n be predicted by use of Bases Figure B 3/4 4-1 andBasestigur$0I3/44-2. The latter figure provides a shift curve YANKEE-ROWE B 3/4 4-/

=. -

i l

}

MAIN COOLANT SYSTEM s

1

}

BASES 1

for the surveiiiance specimens, and calculated shift curves for both the 1/4t and 3/4t locations in the vessel plate. The heatup and cool-l down limit curves in Figures 3.4-2, 3.4-3, 3.4-4 and 3.4-5 have been for core #12. Adjustments adjusted to include the predicted shift in RT forpossibleerrorsinthepressureandtempekuresensinginstruments l

have been included.

l

_ Figures 3.4-2, 3.4-3, 3.4-4 and 3.4-5 shall be revised (without submittal of a change to these Technical Specifications) at each refueling outage by estimating the MWH(t) on the reactor and by indexing through Bases Figures B 3/4.4-1 and B3/4.4-2 to provide shifts in RT)OT at the i

1/4t and 3/4t locations in the reactor vessel beltline plate The heatup 1

and cooldown curves shall be shifted parallel to the temperature axis (horizontal) in the direction of increasing temperature a distance equivalent to the shift in RT at the 1/4t or 3/4t as applicable during the period since the cNes were last constructed. The following e

table provides the appropriate shift parameter to be applied.

I CURVE SHIFT p0SITION

[

Heatup, upper limit 1/4t Heatup, other rate limits 3/4t j

Cooldown, all limits 1/4t

(

The pressure-temperature limit lines shown on Figures 3.4-4 and 3.4-5 for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix j

G to 10 CFR 50.

y L

The requirement that the reactor is not to be made critical out-side the limits specif'ied'provides increased assurance that the proper relationship between main coolant pressure and temperature will be of the Main Coolant System. The maintained relative to the RT)0kave been provided to assure compliance i

limits for reactor criticalit with the minimum temperature requirements of Appendix 'G to 10 CFR 50. -

Heatup to these temperatures will be accomplished by utilizing decay heat and by operating the main coolant pumps.

Temperature requirements for pressurization of the pressurizer-correspond with DTT measured for the material of each component. The L

DTT is defined as the initial Nil Ductility Transition Temperature (NOTT)

I plus 60*F.

A temperature difference of 225'F between the pressurizer and main l

l coolant system is specified to maintain thermal stresses within the surge line below design limits.

l j.

YANKEE-ROWE 8 3/4 4,8 Amendment No. 49 i

-.. - - - ~

CONTAINMENT SYSTEMS BASES 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 stan-dards for the life of the facility. Structural integrity is required to ensure that the containment will withstand the maximum pressure of 31.6 psig in the event of a LOCA. A visual inspection in conjunction with Type A leakage tests is sufficient to demonstrate this capability.

3/4.6.1.7 CONTINUOUS LEAX MONITORING SYSTEM The OPERABILITY of the continuous leak monitoring system provides some assurance that the containment equipment hatch seal, emergency hatch seal, and airlock seal are not leaking excessively between Type A leak tests.

3/4.6.2 CONTAINMENT ISOLATION VALVES The OPERABILITY of the containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in the event of pressurization of the containment. Containment isolation l

within the time limits specified ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a LOCA.

3/4.6.3 COMBUSTIBLE GAS CONTROL The OPERABILITY of the equipment and systems required for the detection and control of. hydrogen, gas ensures that this equipment will be available to maintain ~the hydrogen concentration within containment below its flammable limit during post-LOCA. conditions. The purge system is capable of controlling the expected hydrogen generation associated with 1) zirconium-water reactions, 2) radiolytic decomposition of water and 3) corrosion of metals within containment.

The hydrogen recirculating system is provided to ensure adequate mixing of the containment atmosphere following a LOCA. This mixing action will prevent localized accumulations of hydrogen from exceeding the flammable limit.

YANKEE-ROWE B 3/4 6-3 Amendment No. 49

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FIGURE 5.1-2 i

1 YANKEE-ROWE 5-3 l

i

UESIGN FEATURES CONTROL R00 5.3.2 The reactor core shall contain 24 control rods. At least twenty-two control rods shall contain a nominal 90 inches of absorber material.

The nominal values of this absorber material shall be 80 percent silver, 15 percent indium and 6 percent cadmium. Up to two control rods may have tour 90" blades of Hafnium as the absorber. The 22 silver-indium-cadmium control rods shall be clad with Inconel. The tm Hafnium control rods shall be clad with stainless steel.

5.4 MAIN COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The Main Coolant System is designed and shall be maintained:

a.

In accordance with the code requirements specified in ASME Boiler and Pressure Vessel Code,Section VIII, including all addenda through 1956, and the ANSI (formerly ASI) Standards, Power Piping Code, B31.1,1955 Edition, and 816.5, 1957 Edition, with allowance for nonnal degradation pursuant to the applicable Surveillance Requirements, b.

For a pressure of 2500 psig, and c.

For a temperature of 650*F, except for the pressurizer which is 668*F.

VOLUME 6.4. /. The total water and steam volume of the Main Coolant System is 2940 :ubic feet.

5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown in Figure 5.1-1.

5.6 rUEL STORAGE CRITICALITY l

5.6.1 The new and spent fuel storage racks are designed and shall be maintained with a center-to-center distance between fuel assemblies equivalent te <0.95 with placed in the storage racks to ensure a k the new or spent fuel storage areas flood 88fwith unoorated water. The i

i k

of <0.95 includes a conservative allowance of 3?. ak/k for u8[ertatiities.

l l

TANKEE-ROWE 5-4 l

Amendment No. pf 49 l

i lABLE 6.2-1 MINIMUM SHIFT CREW COMPOSITION #

LICENSE APPLICABLE MODES CATEGORY 1, 2, 3 & 4 5&6 SOL 1

1*

OL 2

1 fion-Licensed 2**

1

Does not include the licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling, supervising Core ALTERNATIONS after the initial fuel loading.

- 0ne additional non-licensed operator is required for MODE 2 except when restarting within four hours of a shutdown for which the cause has been clearly established.

Shift crew composition may be less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate unexpected absence on duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements of Table 6.2-1.

YANKEE-ROWE 6-5 Amendment No. 46

~

l l

ADMINISTRATIVE CONTROLS 6.3 FACILITY STAFF QUALIFICATIONS 6.3.1 Each member of the facility staff listed below shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions.

a.

Plant Superintendent b.

Assistant Plant Superintendnet c.

Chemistry and Health Physics Supervisor d.

Operations Supervisor e.

Reactor Supervisor l

f.

Maintenance Supervisor g.

Instrument and Controls Supervisors h.

Shift Supervisors 1.

Plant Health Fnysicist f

6.4 TRAINING 6.4.1 A retraining and replacement training program for the facility staff shall be maintained under the direction of the Training Coordinator and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and Appendix "A" of 10 CFR Part 55.

6.4.2 A training program for the Fire Brigade shall be maintained under the direction of a member of the plant staff appointed to perform the duties of Fire Protection Coordinator and sha'l meet or exceed the requirements of Section 27 of the NFPA Code-lo?S, except for Fire Brigade training sessions which shall be held at least quarterly.

l 6.5 REVIEW AND AUDIT 6.5.1 PLANT OPERATION REVIEW COMMITTEE i

FUNCTION 6.5.1.1 The Plant Operation Review Comittee (PORC) shall function to.

advise the Plant Superintendent on all matters related to nuclear safety.

(

l I

YANKEE-R0WE 6-6 Amendment Ng. g( 49

r ADMINISTRATIVE CONTROLS COMPOSITION 6.5.1.2 The Plant Operation Review Committee shall be composed of the:

Chairman:

Plant Superintendent Vice Chaiman: Assistant Plant Superintendent Member:

Operations Supervisor Member:

Maintenance Supervisor Member:

Reactor Supervisor l

Member:

Chemistry and Health Physics Supervisor Member:

Instrument and Control Supervisor Member:

Plant Health Physicist ALTERNATES 6.5.1.3 All alternate members shall be appointed in writing by the PORC Chairman to serve on a temporary basis; however, no more than two alternates shall participate as voting members in PORC activities or count toward a PORC quorum at any one time.

MEETING FREQUENCY 6.5.1.4 The PORC shall meet at least once per calendar month and as convened by the PORC Chairman or Vice Chaiman.

QUORUM 6.5.1.5 A quorum of the PORC shall consist of a minimum of five people as follows:

a.

The Chairman or Vice Chairman plus four members, or t.

The Chairman and Vice Chaiman plus three members.

f PESPONSIBILITIES l

6.5.1.6 The Plant Operation Review Committee shall be responsible for:

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

and changes thereto, 2) any other proposed procedures or changes thereto as detemined by the Plant Superintendent j

to affect nuclear safety.

YANKEE-R0WE 6-7 Amendment No. # 49

l ADMINISTRATIVE CONTROLS 1

b.

Review of all proposed tests and experiments that affect-nuclear safety.

c.

Review of all proposed changes to the Technical Specifications.

d.

Review of all proposed changes or modifications to plant systems or u,uipment 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 Manager of Operations and to the Chairman of the Nuclear Safety Audit and Review Comittee.

f.

Review of events requiring 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months written notification to the Commission.

g.

Review of faciiity operations to detect potential nuclear safety hazards, h.

Perfomance of special reviews, investigations of analyses and reports thereon as requested by the Chairman of the Nuclear Safety Audit-and Review Comittee.

1.

Review of the Plant Security Plan and implementing procedures and shall submit recomended changes to the Security Advisor.

j.

Review of the Emergency Plan and implementing procedures and shall submit recomended changes to the Radiation Protection Manager.

l AUTHORITY 6.5.1.7 The Plant Operation Review Comittee shall:

a.

Recomend to the Plant Superintendent, written approval or dis-approval of items considered under 6.5.1.6(a) through (d) above.

I b.

Render determinations in writing with regard to whether or not each item considered under 6.5.1.6(a) through (e) above constitutes an unreviewed safety question.

c.

Provide written notification within ^4 hours to the Manager of Operations of disagreement between the PORC and the Plant Superintendent; however, the Plant Superintendent shall have responsibility for resolution of such disagreements pursuant to 6.1.1 above.

l YANKEE-ROWE 6-8 Amendment No. 46

f ADMINISTRATIVE CONTROLS 1

RECORDS 6.5.1.8 The Plant Operation Review Committee shall maintain written minutes cf each meeting and copies shall be provided to the Manager of Operations and Chairman of the Nuclear Safety Auoit and Review Committee.

6.5.2 NUCLEAR SAFETY AUDIT AND REVIEW COMMITTEE o

FUNCTION 6.5.2.1 The Nuclear Safety Audit and Review (NSAR) Committee shall function to provide independent review and audit of all aspects of plant safety. Adequacy of this review and audit is assured by the cross sec-tion of disciplines required of the Committee membership as described in Section 6.5.2.3.

COMPOSITION 6.5.2.2 The NSAR Committee shall be composed of at least six persons with the Committee membership and its Chairman and Vice Chairman appointed by the Yankee Atomic Electric Company Vice President or such person as he shall designate.

a.

Chairman b.

Vice Chairman Four technically qualified persons who are not members of the c.

plant staff.

QUALIFICATION 6.5.2.3 Membership to the NSAR Committee requires that an individual meet one of the below academic and/or experience requirements:

Bachelor Degree or eauivalent, plus five (5) years total a.

experience in the below listed disciplines, b.

Nine (9) years total experience in the below listed disciplines:

(a) Nuclear Power Plant Technology (b) Reactor Operations (c) Utility Operations (d) Power Plant Design (e) Reactor Engineering (f) Radiation Safety YANKEE-ROWE 6-9 Amendment No. 46

ADMINISTRATIVE CONTROLS (g) Safety Analysis (h) Instrumentation and Control (i) Metallurgy (j) Quality Assurance ALTERNATES 6.5.2.4 All NSAR Comittee alternate members shall be apppointed in writing by the Yankee Atomic Electric Company Vice President, or such person as he may designate, to serve on a temporary basis; however, no more than two alternates shal par *.icipate as voting members in NSAR Comittee activities at any one time.

CONSULTANTS 6.5.2.5 Consultants may be utilized as determined by the NSAR Comittee Chairman to provide expert advice, when needed, to the NSAR Comittee.

1 MEETING FREQUENCY 6.5.2.6 The NSAR Comittee shall meet at least once per six months, I

+ 25%. Special meetings may be held when deemed necessary by Company management or by the Chaiman of the NSAR Committee, or, in the absence of the Chaiman, by the Vice Chaiman.

QUORUM 6.5.2.7 A NSAR Comittee quorum shall consist of a minimum of five members as follows:

The Chaiman er Vice Chaiman plus four members (or two a.

members plus two alternates).

b.

The Chairman and Vice Chairman plus three members ( or one member plus two alternates).

Those personnel from the organization reporting to c.

the Manager of Operations shall always be in the minority.

YANKEE-ROWE 6-10 Amendment No. g 49

l ADMINISTRATIVE CONTROLS (d) Total dissolved gas radioactivity (in curies) and average cor. centration released to the unrestricted area.

(e) Total volume (in liters) of liquid waste released.

(f) Total volume (in liters) of dilution water used prior to release from the restricted area.

(g) Total gross radioactivity (ir curies) by nuclide released based on represent ;ive isotopic analyses performed.

(h) Percent of Technical Specification limit for total radioactivity.

(3) Solid Wastes (a) The total amount of solid waste shipped (in cubic feet).

(b) The totai estimated radioactivity (in curies) involved.

(c) Disposition including date and destination.

6.9.6 Special reports shall be submitted to the Director of the Office of Inspection and Enforcement Regional Office within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification:

a.

Inservice Inspection Program Reviews, Specification 4.4.9.1.

b.

ECCS Actuation, Specifications 3.5.2 and 3.5.3.

c.

Inoperable Meteorological Monitoring Instrumentation, Specifi-cation 3.3.3.3.

d.

Sealed Source leakage in excess of limits, Specification 4.7.6.3.

e.

Radioactive Solid Waste Disposal, Specification 3.7.7.1.

l f.

Fire Detection Instrumentation, Specification 3.3.3.4.

g.

Fire Suppression Systems, Specifications 3.7.10.1, 3.7.10.2 and 3.7.10.3.

h.

Environmental Monitoring Program, Specifications 3.7.11.1,

3. 7.11.2 and 3.7.11.3.

YANKEE-ROWE 6-21 Amendment No. g 49

-_

ML20031C387

Text

Attachment:

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