Text

Amend 65 to License DPR-3,revising Tech Specs to Assure Adherence to Limits & Administrative Requirements Approved Following Implementation of TMI Lessons Learned Category a Items
	ML20003F427
Person / Time
Site:	Yankee Rowe
Issue date:	04/02/1981
From:	Crutchfield D; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20003F422	List: ML20003F421; ML20003F427; ML20003F429; ML20003F430;
References
	NUDOCS 8104210235
	Download: ML20003F427 (26)
	v • d • e

.

UNITED STATES g5 NUCLEAR REGULATORY COMMISSION g

l'(

WASHINGTON, D. C. 20555

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p YANKEE ATOMIC ELECTRIC COMPANY DOCKET NO. 50-29 YANKEE NUCLEAR POWER STAlION (YANKEE-ROWE)

AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 65 License No. DPR-3 1.

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

A.

The application for amendment by Yankee Atomic Electric Company (thelicensee)datedSeptember 16,1980, (Proposed Change No.170) complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

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

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

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

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

810.421o2.35

. ?.

Accordingly, Facility Operating License No. DPR-3 is hereby amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, paragraph 2.C.(2),

and the addition of new paragraphs 2.C.(7) and 2.C.(8), to read as follows:

Technical Specifications (2) The Technical Specifications contained in Appendix A, as revised through Amendment No. 65, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

(7) The licensee shall implement a program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as low as practical levels.

(8) The licensee shall implement a program which will ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions.

3.

This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION hhd.f Dennis M. Crutchfield, ief Operating Reactors Branch #5 Division of Licensing

tachment

2anges to the Technical Speci fica tions 3 ate of Issuance: April 2,1981 l

l O

L

ATTACHMENT TO LICENSE AMENDMENT NO. 65 FACILITY OPERATING LICENSE NO. CPR-3 DOCKET NO. 50-29 Revise Appendix A Technical Specifications by removing the following pages*

and inserting the enclosed pages. The revised pages contain the captioned amendment number and vertical lines indicating the area of change.

Remove Insert IV IV 3/4 3-12 3/4 3-12 3/4 3-12A 3/4 3-14 3/4 3-14 3/4 3-15 3/4 3-15 3/4 3-15A 3/4 3-29 3/4 3-30 3/4 3-31 3/4 4-6 3/4 4-6 3/4 4-6A 3/4 4-7 3/4 4-7 B3/4 3-2 B3/4 3-2 B3/4 4-2 B3/4 4-2 6-5 6-5 6-6 6-6 i

0verleaf pages III, 3/4 3-11, 3/4 3/13, 3/4 3-16, 3/4 4-8, B3/4 3-1, and B3/4 4-1 are included for document completeness.

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIRENENTS SECTION Page 3/4.0 APPLI'CABILITY.............................................

3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 B0 RATION CONTROL 3/4 1-1 Shutdown Margin - T,yg > 330*F..........................

Shutdown Margin - T

< 330*F...~...m.m..... m m.m.

3/4 1-3 avg Boron 0ilution.........................................

3/4 1-5 Moderator Temperature Coefficient......................

3/4 1-7 3/4.1.2 30 RATION SYSTEMS Flcw Paths - Refueling.................................

3/4 1-8 Flow Paths - Shutdown..................................

3/4 1-9 Flow Paths - Operating.................................

3/4 1-11 f

Charging Pumps - Refueling.............................

3/4 1-13 Charging Pump - Shutdovn...............................

3/4 1-14 C ha rg i ng Pump s - Op e ra ti ng....,.........................

3/4 1-15 Boric Acid Mix Tank Gravity Fe'ed Connecticn - Shutdown and Refueling........................................

3/4 1-16 Boric Acid Mix Tank Gravity Feed Connection -

0cerating............................................

3/4 1-17 i

Borated Water Sources - Refueling......................

3/4 1-18 Barated Water Scurces - Shutdown.......................

3/4 1-19 Scrated Water Sources - Operating......................

3/4 1-21 1

3/4.1.3 MOVABLE CONTROL RODS j

Control Rod Operability................................

3/4 1-23 Position Indicator Channels............................

3/4 1-25 Rod Drop Time..........................................

3/4 1-26 Shutdown Red Insartion Limi...........................

3/4 1-27 i

C ntrol Red Insertien Limits...........................

3/4 1-25 fVAMEE-RCWE III-

INDEX LIMITI!C CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS S ECTION j[ajgt 3 /4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 PEAK LINEAR HEAT GENERATION RATE................... 3/42-1 3 /4.2.2 NUCLEAR HEAT FLUX HOT CHANNEL FACTOR............... 3/42-7 3/4.2.3 NUCLEAR ENIHALPY RISE HOT CHANNEL FACIM........... 3/42-9 3 /4.2.4 DNB PARAMETERS.....................................

3/4 2-11 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTIVE SYSTEM INSTRUMENTATION.......... 3/43-1 3 /4.3.2 ENGINEERED SAFEGUARDS SYSTEM INSTRUMENTATION....... 3/4 3-11 3/4.3.3 MONITORI!G INSTRUMENTATION Radiation Monitori ng Ins trumentation............... 3/4 3-17 Incore Detection System............................

3/4 3-23 Me t e or ological Ins trume nt ation..................... 3/4 3-24 Fir e De t ect ion Ins trume nt a tion..................... 3/4 3-27 Accident Monitoring Instrumentation................

3/4 3-29 3 /4.4 MAIN COOLANT SYSTEM 3 /4.4.1 MAIN COOLANT LOOPS N o rm a l O pe ra t i o n................................... 3/44-1 l

Isolated Loop......................................

3/44-3 Main C oola nt Loop S t artu p.......................... 3/44-4 3 /4.4.2 S A FE TY VALVE S - SHU TD0WN........................... 3/44-5*

3 /4.4. 3 S A FE TY VALVE S - OPE RATI N3.......................... 3/44-6 3 /4.4.4 PRESSURIZER........................................

3/44-7 3 /4.4.5 MAIN COOLANT SYSTEM LEAKAGE Leakage Detection Systems..........................

3/44-8 Operational Leakage................................

3/4 4-10

i'ith 3/4 4-Sa.

YANKEE-ROWE IV Amendment No.44T, 56, 65 I

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1

.7.1~.7"~

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h:;:;!:;::

~~

l !'i!!E.' MENTATION _

l,?j[3 IM.3.2 ENGINEERED SAFEGUARDS SYSTEM INSTRUMENTATION L:!'IT:N3 CONDITION FOR OPERATION

!==e=

EN#'i' =

hP~'

=

3.3.2 The Engineered Safeguards System (ESS) instrumentation channels ian: sensors shown in Table 3.3-2 shall be OPERABLE with their trip set-E..

c'nts set consistent with the values shown in the Trip Setpoint column ll. """

cf Table 3.3-3.

is....__

A?:LICA31LITY: As shown in Table 3.3-2.

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y=m.==

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a.

h*ith an ESS instrumentation channel senser trip setpoint less con-servative than the value shown in the Trip Setpoint column of Table 3.3-3, declare the channel inoperable and apply the applicable

"@=-

ACTION requirement of Table 3.3-2 until the channel is restored to OPERABLE status with the sensor trip setpoint adjusted consistent

n

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with the Trip Setpoint value.

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t.

aith an ESS instrumentation channel inoperable, take the ACTION shown in Table 3.3-2.

s

,E P.'E:LLA'1CE REOUIREMENTS

...=

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1 Each ESS instrumentation channel and sensor shall be demon-ls.3.2.1t:ed 0?ERABLE by the performance of the CHANNEL CHECX, CHANNEL

~

j'M*.:i:.ATI0ft and CHANNEL FUNCTIONAL TEST operations during the OPERATIONAL

ES ar.d at the frequencies shown in Table 4.3-2.

i. 3.2.2 The total bypass function of all bypasses shall be demonstrated t

,::EF.* 5LE at least once per 18 months during CHANNEL CALIBRATION testing

':f at:h channel affected by bypass operation.

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.v.EE :.CWE 3/4 3-11

P TABLE 3.3-2 2

5 7

ENGINEERED SAECUARDS SYSTEM INSTRUMENTATION

$g MINIMUM TOTAL NO.

CllANNELS OlANNELS OF CilANNELS AND SENSORS AND SENSORS APPLICABLE FUNCTIONAL UNIT AND SENSORS TD TRIP OPERABLE MODES ACTION 1.

SAETY INJECTION a.

Actuation Channel #1 1

1 1

1, 2, 3(2)(3) 10

1) RPS Low Main Coolant Pressure Channel 1

1 1

1, 2, 3(2)(3) 10

2) liigh Containment w

Pressure Sensor 1

1 1

1, 2, 3(2)(3) 10 5

w

3) Manual Initiation 1

1 1

1,2,3,4, SCI) 10 b.

Actuation Channel #2 1

1 1

1, 2, 3(2) 10

1) Low Main Coolant Pressure Sensor 1

1 1

1, 2, 3(2) 10

2) liigh Containment Pressure Sensor 1

1 1

1, 2, 3(2) 10

3) Manual Initiation 1

1 1

1, 2, 3, 4, 5(1) 10 2.

CONTAINMENT ISOLATION a.

Manual Initiation 2

1 1

1, 2, 3, 4, 5(1) 10 p$

1 b.

Actuation Channel A 1

1 1

1, 2, 3, 4, 5(1) 10 h

1) liigh Containment m

S Pressure Sensor 1

1 1

1, 2, 3, 4, 5(1) 10 v.

?

2) Safety Injection (All Safety Injection Initiating Functions and Requirements)

=

4 TA n t.It.l. l - 2 (Ces se t t ini...l )

(j El ENt;l NI:l:K i th; !;Al I!Gtl Altt):3 :;Y!!TI H I N: THtf MENTATI ON I.i MINIMUM TOTAL NO.

CilANNELS CilANNELS OF CilANNELS AND SENSORS AND SENSORS APPLICABLE FUNCTIONAL UNIT AND SENSORS TO TRIP OPERABLE MODES ACTION 2.

CONTAINMENT ISOLATION (Continued) 1, 2, 3, 4, 5(1) 10

c. Actuation Channel B 1

1 1

1) High Containment CI) 10 Pressure Sensor 1

1 1

1,2,3,4,S

2) Safety Injection

( All Safety Injection Initiating Functions and Requireneents)

R>

0 y.

r ii.

9e

TABLE 3.3-2 (Continued)

TABLE NOTATION II) Trip function may be bypassed in this MODE with main coolant

.g pressure < 300 PSIG.

(2) Trip function may be bypassed in this MODE with main coolant pressure

< 1800 PSIG.

5.h" (3) Automatic initiation of Actuation Channel !1 may ba hypassed in this

=:

BODE during functional test of the !iain Coolant System pressure channel.

ACTION STATEMENTS ACTION 10 - With the number of OPERABLE channels or sensors one less than the Total Number of Channels or sensors, 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 SHUTriWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ; however, one safety inject n channel high containment pressure sensor may be bypassed f up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.2.1.

~

S YANKIE-ROWE 3/4 3-13

TABLE 3.3-3 ENGINEERED SAFEGUARDS SYSTEM INSTRUMENTATION TRIP SETPOINTS

.ra;CMONAL UNIT TRIP SETPOINT 1.

SAFETt INJECTION a.

Actuation Channel #1

1) RPS Low Main Coolant Pressure Channel

);1700 psig

2) High Containment Pressure Sensor i 5 psig

3) Manual Initiation Not Applicable b.

Actuation Channel #2

1) Low Main Coolant Pressure Sensor

>;1700 psig

2) High Containment Pressure Sens or 1 5 psig

3) Manual Initiation Not Applicable 2.

CONTAINMENT ISOLATION a.

Manual Initiation Not Applicable b.

Actuation Channel A

1) High Containment Pressure Sens or i 5 psig

2) Safety Injection (All Safety Injection Setpoints) c.

Actuation Channel B

1) High Containment Pressure j[5 psig Sensor

2) Safety Injection (All Safety Injection Setpoints) l

' Amendment No. 347 65 YASKEE-R3WE 3/4 3-14

TABLE 4.3-2 I"

E!CINEERED SAFEGUARDS SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS b

CHANNEL MODES IN WHICH m

CilANNEL CilANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CilECK CALIBRATION TEST REQUIRED 1.

SAFEIY IRJECTION a.

Actuation Channel #1 S

N.A.

M(1) 1, 2, 3#

1) RPS Low Main Coolant Pressure Channel S

R(3)

M(2) 1, 2, 3#

2) liigh Containment Pressure Sensor S

R(3)

M(3) 1, 2, 3#

3) Manual Initiation N.A.

N.A.

R 1,2,3,4,5*

M b.

Actuation Channel #2 S

N.A.

M(1) 1, 2, 38 y

1) Low Main Coolant C;

Pressure Sensor S

R(3)

M(3) 1, 2, 3#

2) liigh Containment Pressure Sensor S

R(3)

M(3) 1, 2, 38

3) Manual Initiation N.A.

N.A.

R 1,2,3,4,5*

2.

CONTAINMENT ISOLATION a.

Manual Initiation N.A.

N.A.

R 1, 2, 3, 4, 5*

b.

Actuation Channel A S

N.A.

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

k

1) liigh Containment g,

Pressure Sensor S

R(3)

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

?

E

2) Safety Injection (All Saf ety Injection Surveillance Requirements)

E.

=

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TAlu.n 4.

.' i c.... i.....I )

I,'I, ENGINEEl:ED SAFECII Alt!)S SYSTEM TNSTRifMENTATTON p,

SilitVEII I.ANCE IIEqllll(19 TENTS p!

n CHANNEL MODES IN WHICH CllANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CllECK CALIBRATION TEST REQUIRED 2.

CONTAINMENT ISOLATION (Continued) i

c. Actuation Channel B 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*

2) Safety Injection (All Safety Injection Surveillance Requirements) 4

.f 0

4 7

A 1

ait O

ct O

i 1

4 m,.

.= =

=

'.'.? Y,.

($5.f..fb...[

TABLE 4.3-2 (Continued)

EE"?52 i TABLE NOTATION

.).g

=

(1) When shutdown with main coolant pressure < 1000 psig, if not

.. = =,

performed within the previous 31 days, p.:j:.=3

~

.=g=.

(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

i=

iiip=E previous 31 days.

....:?-

~

I (3) The test shall include exercising the sensor by applying either a

'Eni"-

vacuum or pressure to the appropriate side of the sensor.

jy]

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

~= ~l.'

d:.=.

Not required in this MODE with main coolant pressure < 300 PSIG.

]

Not required in this MODE with main coolant pressure < 1700 PSIG.

... }[.! '

.:g =:

.. :.. :_ f.. ' *-*

=

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T YMKEE-ROWE 3/4 3-16 Amendnent No. 49

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4 l I INSTR *.* MENTATION l

AOCIDENT MONITORING INSTRUMENTATION 1

LIMITING CONDITION FOR OPERATION

,3.3.3.5 The accident monitoring instrumentation channels shown in Table 3.3-7 shall be OPERABLE.

A?PLICABILITY: MODES 1, 2, and 3.

s ACTION:

With the number of OPERABLE accident monitoring instrumentation channels a.

less than the Total Number of Channels shown in Table 3.3-7, either restore the inoperable channel (s) to OPERABLE status within 7 days, or 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 .

b.

With the number of OPERABLE accident monitoring instrumentation channels less than the MINIMUM CRANNELS OPERABLE requirements of Table 3.3-7, either restore the inoperable channel (s) to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT 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.

c.

S'.*FXEILLANCE REQUIREMENTS h

,4.3.3.6 Each accident monitoring instrumentation channel shall be deconstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL l C ALIBRATION operations at the frequencies shown in Table 4.3-5.

A l

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IYANKEE-ROWE 3/4 3-29 Amendment No. 65

(!

l TABLE 3.3-7 ACCIDENT MONITORIE INSTRUMENTATION MTAL NO.

MINIMUM CH8.~iMELS INSTRUMENT OF CHANNELS OPERAIE 1.

Pressurizer Water Level 2

1 2.. Auxiliary Feedwater Flow Rate 4

3

.i. Main Coolant System Subcooling 1

1 Margin Monitor 4.

PORV Position Indicator (Acoustic 1

1 Accelerometer) 5.

PORV Position Indicator (Temperature 1

0 Indication, RID) 6.

PORV Block Valve Position 1

1 Indicator 7.

Safety Valve Position Indicator 1

1 (Acoustic Accelerometer) 8.

Safety Valve Position Indicator 1

0 (Temperature Indication, RE) t Amendment No. 65 YANKEE-ROWE 3/4 3-30

l TABLE 4.3-5 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL INSTRUMENT CHECK CALIBRATION _

l.

Pressurizer Water Level M

R 2.

Auxiliary Feedwater Flow Rate M

R

."3.

Main Coolant System Subcooling M

R Margin Monitor 4

PORV Position Indicator (Acoustic M

R Accelerometer) 5.

PORV Position Indicator (Temperature M

R f

Indication, RTD)

'i 6.

PORV Block Valve Position M

R Indicator 7.

Safety Valve Position Indicator M

R (Acoustic Accelerometer)

E.

Safety Valve Position Indicator M

R (Temperatura Indication, RTD) s a

A, t

Amendment No. 65 YANKEE-ROWE 3/4 3-31

MAIN COOLANT SYSTEM S AFETf VALVES - OPERATI)G LIMITI)G CONDITION FOR OPERATION 3.4.3 At least the following safety and/or relief valves shall be OPERABLE:

Two pressurizer code safety valves shall be OPERABLE with lift a.

settings of 2485 psig +0,

-3% and 2560 poig +0, -3%, respectively.

b.

The power operated relief valve PR-SOV-90 and its associated block valve PR-MOV-512 shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

With one pressurizer code safety valve inoperable, either restore a.

the inoperable valve to OPERABLE status within 15 minutes or be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.

With PR-SOV-90 or PR-MOV-512 inoperable:

1.

With the PORV inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore the l

PORV to OPERABLE status or close the associated block valve and remove power f rom the block valve; 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 .

2.

With the PORV's block valve inoperable, within I hour either restore the block valve to OPERABLE status or close the block valve and remove power from the block valve;. 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 .

i 1

I YANKEE-ROWE 3/44-6 Amendment No. 65

.9 SL?XEILLM;CE REQU1nEMENTS i

4.4.3.1 Each preasurizer code safety valve shall be demonstrated OPERABLE lwithliftsettingsof2485psig+0,-3%and2560psig+0,-3%,respectively, in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, 1974 Edition, and Addenda through Summer 1975.

4.4.3.2 The PORV chall be demonstrated OPERABLE:

By performance of a CRANNEL FUNCTIONAL TEST, excluding valve a.

operation, v'-en shutdown for longer than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months if not performed in the pre s 31 days, and b.

At least or..

per 18 months by performance of a CHANNEL r

CALIBRATION.

I 4.4.3.3 The block valve shall be demonstrated OPERABLE at least once per

92 days by operating the valve through one complete cycle of full travel.

i I

4.4.3.4 The emergency power supply for the block valve shall be demonstrated i

OPERA 3LE at least once per 18 months by transf erring motive and control power from the normal to the emergency power supply and operating the valve through a complete cycle of full travel.

l 1

d l

l.,

l YANKEE-RO'JE 3/4 4-6A Amendment No. 65

MAIN COOLANT SYSTEM PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.4 The pressurizer shall be OPERABLE with a bubble, at least 37.5 kw of pressurizer heaters, and a water level between 80 and 150 inches.

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

ACTION:

a.

In M3DE 1 or 2 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 .

In MODE 4* with the pressurizgr bubble volume < 198 f t3, restore b.

the bubble volume to 2; 198 f t within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

With the pressurizer inoperable due to an inoperable emergency c.

power supply to the pressurizer heaters, either restore the inoperable emergency power supply within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT SIANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN 3

within the folleving 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

l SURVEILLANCE REQUIREMENTS i

4.4.4.1 The pressurizer shall be demonstrated operable by:

The pressurizer water volume shall be determined to be within a.

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

b.

The emergency power supply for the pressurizer heaters shall be demonstrated OPERABLE at least once per 18 months by transferring power from the normal to the emergency power supply and encrgizing the heaters.

4.4.4.2 In M3DE 4* the pressurizer level:

4 Shall be verified to be < 129", equivalent to a pressurizer bubble a.

volume ]; 198 f t3, at least once per 30 minutes.

I b.

Channel (s) shall be demonstrated OPERABLE by performance of a:

1.

CRANNEL CHECK at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , 2.

CHANNEL FUNCTIONAL TEST when shutdown longer than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months if not performed within the previous 31 days, and

(

l 3.

CRANNEL CALIBRATION at least once per 18 months.

henever MOS temperature is between 300 F and 324 F.

0 W

i YANKEE-ROWE 3/44-7 Amendment No. 49,,50, 59', 65

_ ~ _.

. ~. _. _

i

~?

,M IN COOLANT SYSTEM

?.Y 3 '4.4.5 MAIN COOLANT SYSTEM LEAKAGE L E. JASE DETECTION SYSTEMS _

l

' L:M: TING" CONDITION FOR OPERATION The following Main Coolaint System leakage detection systems i

~ 3. 4. 5.1

, snail be OPERABLE:

At least one containment atmosphere particulate radioactivity l

i a.

monitoring system, j

The containment drain tank level monitoring system.

b.

The incore detection system thimble leak alarm system.

c.

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

.CTICN:

u With the above required radioactivity monitoring leakage detection

!, z.

sys:tc. inoperable, operation may continue for up to 7 days provided:

h Main Coolant System water inventory balance is performed at i

=

1.

least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

The other above required leakage detection systems are OPERABLE, 2.

and Appropriate grab san.ples are obtained and analyzed at least 3.

\\

once per hour:

ctherwise, be in at least HOT STANDSY 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 .

With the containment drain tank level monitoring system inoperable, restcre the inoperable system to OPERASLE status within 7 days or b.

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 SHL'TOOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months , With the incore detection system thimble leak alarm system inoper-able, restore the leak alarm system to OPE?A5LE status within 7 days c.

cr close all thimble isolation valves; restore the leak alarm system to OPERASLE status within 31 days or be in fcil: wing 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months , Amendment No. 58 3/4 4-8 j12.EEE-RCWE

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3/4.3 INSTRUMENTATION

. BASES 3/4.3.1 and 3/4.3.2 REACTOR PROTECTIVE SYSTEM (RPS) AND _ ENGINEERED SAFEGUARDS SYSTEM (ESS) INSTRUMENTATION The OPERABILITY of the RFS and ESS instrumentation systems and permissive circuits ensure that 1) the associated ESS action and/or RPS trip will be initiated when the parameter monitored by each channel or combination thereof exceeds its -setpoint, 2) the specified coinci-dence logic is maintained, 3) sufficient redundancy is maintained, and

4) sufficient system functional capability is available for RPS and ESS purposes from diverse parameters.

The OPERABILITY of these systems is required to provide the overall reliability, redundance and diversity assumed available in the facility design for the protection and mitigation of accident and transient con-ditions. The integrated operation of each of these systems is consistent with the assumptions used in the accident analyses.

The surveillance requirements specified for these systems ensure that the overall system functional capability is maintained comparable to the original design standards. The periodic surveillance tests -

performed at the minimum frequencies are sufficient to demonstrate this capability.

3/4.3.3 MONITORING INSTRUMENTATION 3/ 4. 3. 3.1 RADIATION MONITORING INSTRUMENTATION

- The OPERABILITY of the radiation monitoring channels ensures that

1) the radiation levels are continually measured in the areas served by the individual channels and 2) the alarm or automatic action is in-itiated when the radiation level trip retpoint is exceeded.

YANKEE-ROWE B 3/4 3-1 Amendment No. 46, o

%g

,'s g 1

, INSTRUMENY.ATION

! BASES 3 /4. 3. 3. 2 INCORE DETECTION SYSTEM The OPERABILITY of the incore detection system with the specified cinimun comple=ent of equipment ensures that the measurements obtained from use of this system accurately represent the reactor core power distribution.

The OPERABILITY of the neutron detectors is demonstrated prior to use by

~

irradiating each detector to be used and normalizing their outputs.

3 /4. 3. 3. 3 METEOROLOGICAL INSTRUMENTATION Th-OPERABILITY of the meteorological instrumentation ensures that sufficient meteorological data is available for estimating potential

! radiation doses to the public as a result of routine or accidental release of radioactive materials to the atmosphere. This capability is required to evaluate the need for initiating protective measures to protect the health l and safety of the public (and is consistent with the recommendations of I Regulatory Guide 1.23, "Onsite Meteorological Programs").

i 3 /4. 3. 3. 4 FIRE DETECTION INSTRUMENTATION OPERABILITY of the fire detection instrumentation ensures that

! adequate warning capability is available for the prompt detection of fires.

This capability is required in order to detect and locate fires in their early stages. Prompt detection of fires will reduce the potential for damage to safety related equipment and is an integral element in the overall

~

facility fire protection program.

i In the event that a portion of the fire detection instrumentation-l 1s inoperable, the establishment of freqdent fire patrols in the affected areas is required to provide detection capability until the inoperable instrumentation is restored to OPERABILITY.

f I

3 /4. 3. 3. 5 ACCIDENT MONITORING INSTRUMENTATiuN i

.The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to l

> monitor and assess these variables during and following an accident. This

{

capability is consistent with the recommendations of Regulatory Guide 1.97,

" Instrumentation for Light-Water-Cooled "uclear Power Plants to Assess Plant Conditions During and Following an Accidcnt," December 1975 and NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations."

l i

l l

i.

' l YANKEE-ROWE B 3/4 3-2 Amendment No. 34I 65

4s=-

Su 3/4.4 MAIN COOLANT SYSTEM BASES 3 /4.'4.1 MAIN COOLANT LOOPS The plant is designed to operate with all main coolant loops in operation, and maintain DNBR above 1.30 during all normal operations and anticipated transients. With 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 locps will cause a reactor trip. A loss of flow or low SG water level 1

7 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 Cocling 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 ret.

The requirement to maintain the boron concentration of an isolated loop greater than or equal to the boron concentration of the operating loops ensures that no reactivity addition 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 prior to loop startup prevents any power spike which could result from this cool water induced reactivity transient.

The prohibition on starting a main ccolant purp witnout a bubble in the pressurizer or with a SG/MCS terperature difference of > 100*F requirement is necessary to prevent exceeding the isotherral Appendix G curve limits under the most restrictive MC flow initiation transient'if the PORV or SCS relief valves fail.

3/a.a.2 and 3/4.4.3 SAFETY VALVES The pressurizer code safety valves operate to prevent the Main Ccolant System from being pressurized above its Safety Limit of 2735 psig.

Each safety valve is designed to relieve 92,000 lbs per hour of saturated steam at the valve set point.

Th-relief capacity of a single safety valve is adequate to relieve any cverpressure ccndition which cculd occur during shutdcwn.

In the ev n: that no safety valves are C?E'lABLE, an ccerating Shutdctn Ccoling System connected to the Main

-oc,lant Syster prcvides overpressure relief capability and will prevent Main Coolant System overpressurizaticn during shutdcwn.

YANKEE-ROWE B 3/4 4-1 Amendment No.19, 59 sEP 14 19

-. rXN COOLANT SYSTEM 3ASES t

SAE Y VALVES (Continued)

During operation, all pressurizer code safety valves must be OPERABLE

o prevent the Main Coolant System from being pressurized above its safety 11:10 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 setpoint is reached (i.e., no credit is taken for a direct reactor i 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.

1 The MCS passive code safety valves, SV-204 and SV-205, are available l

0 to relieve expansion or mass addition at all HCS temperatures < 300 F.

Ea:h of the two valves has the capacity to relieve 101 gpm of 400 F water at a set pressure of 47.5 psig plus 10% accumulation. The low setpoint PORV,

?R-S3V-90, is available to relieve expansion or mass addition at MCS temperature _< 324 F.

The power operated relief valve (PORV) operates to relieve MCS

{ pres sure below the setting of the pressurizer code safety valves. This relief valve has a remotely operated block valve to provide a positive shatof f capability should the relief valve become inoperable. The electrical j pcwer for the block valve is capable of being supplied from an emergency

' p ver source to ensure the ability to seal this possible MCS leakage path.

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3 '4.4.4 PRESSURIZER A bubble in the pressurizer ensores hat the Main Coolant System a hydraulically solid system and is capable of accocmodating pressure l

1s not surges during operation. The b2bble also protects the pressurizer code safety valves and power operated relief valve against water relief. The j ' p:wer operated relief valve and bubble function to relieve Main Coolant i' System pressure during all design transients up to and including the design l ' step load decrease with steam dump. Operation of the power operated relief the undesirable opening of the spring-loaded pressurizer

!l valve minimizes c :de saf,ety valves.

3 The pressurizer surge volume of > 198 f t requirement pr ovide s sufficient tice for corrective operator action to prevent exceeding the

! isothermal Appendix G curve limits under the most restrictive mass input l

ia:cident should the PORV f ail to operate at.the low setpoint.

l The requirement that 37.5 kw of pressurizer heaters 'and their associated contrcis be capable of being supplied electrical power f rom an

, e:ergency bus provides assurance that these heaters can be energized during a loss of of fsite power condition to maintain natural circulation at HOT i

I S IAN33Y.

YANT.EE-ROWE B 3/4 4-2 Amendment No. 59, 65 i

s

,a s - e

4 TABLE 6.2-1 l

MINIMUM SHIFT CREW COMPOSITION i LICENSE CATECORY APPLICABLE HODES QUALIFICATIONS 1, 2, 3 & 4 5&6 SRO 1

1*

RO

'2 1

Non-Licensed Auxiliary 2**

1 Operator Shift Technical Advisor 1

None Required Does not include the licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling, supervising CORE ALTERATIONS af ter the initial fuel loading.

- One 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.

I 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.

1 4

YANKEE-ROWE 6-5 Amendment No. Jg5 65

ll ADMIN 1STRATIUECONTROLS i

5 6.3 FACILITY STAFF QUALIFICATIONS 5.3.1 Each member of the facility staff listed below shall meet or exceed the minimum qualifications of ANS1 N18.1-1971 for comparable positions, except for the Shif t Technical Advisor who shall have a bachelor's degree Or equivalent in a scientific or engineering discipline with specific training in plant design, and response and analysis of the plant for

ransients and accidents.

'a.

Plant Superintendent b.

Assistant Plant Superintendent Chemistry and Health Physics Supervisor i

d.

Operations Supervisor e.

Reactor Supervisor f.

Maintenance Supervisor g.

Instrument and Controls Supervisors h.

Shift Supervisors 1.

Plant Health Physicist j.

Shift Technical Advisor 5.4 TRAINING lshallbemaintainedunderthedirectionof'gprogramforthefacilitystaff

,5.4.1 A retraining and replacement trainir.

the Training Coordinator and

,shall ceet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and Appendix "A" of 10 CFR Part 55.

l6.4.2 A training program for the Fire Brigade shall be maintained 'undes

he direction of a member of the plant staf f appointed to perform the duties af Fire Protection Coordinator and shall meet or exceed the requirements of Section 27 of the NFPA Code-1976, except for Fire Brigade training se ssions which shall be held at least quarterly.

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's.5 REVIEb' AND AUDIT 6.5.1 PLANT OPERATION REVIEW COMMITTEE FUNCTION l 5.5.1.1 The Plant Operation Review Committee (PORC) shall function to advise

ht Plant Superintendent on all matters related to nuclear safety.

YANKIE-ROWE 6-6 Amendment No. 49' 65

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ML20003F427

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