ML20003E444

Proposed Changes to Tech Specs Re Refueling Reactor for Core Xv Operation by Replacing 36 Fuel Assemblies.Includes Pages VI,1-6,2-2,2-3,2-6,B2-4,B2-5,3/4 1-1,3/4 1-29,3/4 2-4 Through 2-7,3/4 3-2,3/4 3-3,3/4 3-8,3/4 3-9 & 3/4 3-12

ML20003E444
Person / Time
Site:	Yankee Rowe
Issue date:	03/26/1981
From:	YANKEE ATOMIC ELECTRIC CO.
To:
Shared Package
ML20003E443	List: ML20003E442 ML20003E444 ML20003E447
References
NUDOCS 8104030478
Download: ML20003E444 (35)

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ATTACID!ENT A PROPOSED TECHNICAL SPECIFICATION CHANGES l

l 810 4 0 3 0Y7h

SUMMARY

OF 1ECHNICAL SPECIFICATION CHANCES Delete the Insert the Following Following Pages Revised Pages VI VI 1-6 1-6 2-2 2-2 2-3 2-3 2-6 2-6 B2-4 B2-4 B2-5 B2-5 3/4 1-1 3/4 1-1 3/4 1-29 3/4 1-29

-3/4 2-4 3/4 2-4 3/4 2-5 3/4 2-5 3/4 2 3/4 2-6 3/4 2-7 3/4 2-7 3/4 3-2 3/4 3-2 3/4 3-3 3/4 3-3 3/4 3-8 3/4 3-8 3/4 3-9 3/4 3-9 3/4 3-12 3/4 3-12 3/4 3-13 3/4 3-13 3/4 3-14 3/4 3-14 3/4 3 3/4 3-15 3/4 6-12 3/4 6-12 3/4 6-14 3/4 6-14 3/4 6-14a (new)

-3/4 7-5 3/4 7-5 3/4 7-9 3/4 7-9 53/4 1-1

. B3/4 1-1 B3/4 7-3 B3/4 7-3 5-1 5-1 t

t N

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INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS PAGE SECTION 3/4.6.3 COMBUSTIBLE CAS CONTROL 3/4 6-16 Hydrogen Analyzer....................................

3/4 6-17 Hydrogen Vent System.................................

Atmosphere Recirculation System......................

3/4 6-18 3 /4.7 PLANT SYSTEMS 3 /4.7.1 TURBINE CYCLE 3/4 7-1 Safety Va1ves........................................

3/4 7-5 E me rge nc y F e e d wa t e r Sy s t em...........................

Primary and Demineralized Water Storage Tanks........ 3/47-6 3/4 7-7 Activity.............................................

Main Steam Non-Return Va1ves.........................

3/4 7-9 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION......

3/4 7-13 3/4.7.3 PRIMARY PUMP SEAL WATER SYSTEM (Dele ted)............. 3/4 7-14 3/4.7.4 SERVICE WATER SYSTEM (Dele ted)....................... 3/4 7-16 l

3 /4.7. 5 CONTROL ROOM VENTILATION SYSTEM EMERGENCY SHUTDOWN... 3/4 7-18 3/4.7.6 SEALED SOURCE CONTAMINATION..........................

3/4 7-19 3/4.7.7 WASTE EFFLUENTS Radioactive Solid Waste..............................

3/4 7-21 l

3/4 7-22 Radioac t ive Liquid Wa st e.............................

3/4 7-23 l

Radioactive Gaseous Waste............................

I l

3/4.7.8 E NVIRONMENTAL MONITORING............................. 3/4 7-24

{

3 /4.7.9 SHOCK SUPPRESSORS (S NUBBERS )......................... 3/4 7-27 3/4 7-30 3/4.7.10 FIRE SUPPRESSION SYSTEMS.............................

l 3/4 7-37 l

3/4.7.11 PENETRATION FIRE BARRIERS............................

YI YANKEE-ROWE

a TABLE 1.1 OPERATIONAL MODES REACTIVITY Z RATED AVERAGE *COOLAhT MODE CONDITION, K,fg THERMAL POWER

• TEMPERATURE __

l.

POWER OPERATION 10.99

> 2%

1 330 F 2.

STARTUP 1 0.99 12%

1 330 F 0

3.

HOT STANDBY

< 0.99 0

> 330 F 0

4.

HOT SHUTDOWN

< 0.96 0

330 F > T,yg > 200 F 0

5.

COLD SHUTDOWN

< 0.96 0

1 200 F 0

6.

REFUELING **

10.95 0

i 140 F

• Excluding decay heat.

• • Reactor vessel head unbolted or removed and fuel in the vessel.

YANKEE-ROWE l-6

660.

640 ta.

620 x

.'~

~ ~

v

~

up MAIN COOLANT A.

t A-SYSTEM c-600 PRESSUFI g

-m.

K_

~ *-.

N

'-x__

j

'__ ~

__~

~_

3 580

~' '_' --'_

-=:

m x

" 2600 psia ou

'__ -~

E

m i_

x 2400 psia x*

x x'_--

560

~.

u

,.'x

's

-2200 psia x

E

-K u

x_

c -

-540

~

x 7

-2000 psia z

~

'x' 520

-- -w I800 psia 3 1600 psia 500 70 80-90 100 110 120-130 Indicated Reactor Power, Percent REACTOR CORE SAFETY LIMIT - ALL LOOPS IN OPERATION FIGURE 2.1-3 YANKEE-ROWE 2-2

660 I

640-A u.-

i'-%

o' 620 k

3

'~_

_ _ '-'-=

SYSTEM O

'.'-=__

600 MAIN COOLANT E.

g

~ _ _

PRESSURE x

x g

N 1

x x

'x a

g __

x 580

2600 psla

.e v

'(

~_'

's_

g g

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-' 2400 psia 560 v>

..'m

'm r-

'N-.;

'2200 psia O

',_~--

O 5zo

.c

- '~.

y

~

~_.

4

'2000 psia s

~_

~_

~

520 x

m.

x i1800 psia

.-~.

-_+ _

x_

500

.1600 psia 50 60 70 80 90 100 110 Indicated Reactor Power, Percent REACTOR CORE SAFETY LIMIT - 3 LOOPS IN OPERATION FIGURE 2.1-2 YANKEE-ROWE 2-3

d TABLE 2.2-1 (continued)

REACTOR PROTECTIVE SYSTEM INSTRUMENTATION TRIP SETPOINTS e

N FUNCTIONAL UNIT TRIP SETPOINT 8.

High Hain Coolant System Pressure

< 2200 psig l

9.

Low Main Coolant System Pressure

> 1800 psig 10.

High Pressurizer Water Level

< 200 inches 11.

Low Steam Generator Water

> - 13*

12.

Turbine Trip Not Applicable

13. Generator Trip Not Applicable i '

Y*

14.

Main Steam Isolation Trip Logic

> 200 psig

• Where 0 inches corresponds to 10" above the feed ring centerline.

1 4

LD1ITING SAFETY SYSTEM SETTINGS BASES Low Main Coolant Flow (Steam Generator AP)

The Low Main Coolant Flow trips provide core protection in the event of a loss of one or more main coolant pumps.

Above a power of 15 MWE, with 4 main coolant pumps operating, an automatic reactor trip will occur if the flow in any two loops drops below 80% of nominal full loop flow and, with 3 main coolant pumps operating, automatic reactor trip will occur if the flow in any single operating loop drops below 80% of nominal full loop flow. The setpoints specified are consistent with the value essumed in the accident analysis.

Low Main Coolant Flow (Main Coolant Pump Current)

The Low Main Coolant Flow trips provide core protection in the event of a loss of one or more main coolant pumps.

Above a power of 15 MWE, with 4 main coolant put ps operating, an automatic trip will occur if the main coolant pump motor current is outside the limits on any two pumps, and with 3 main coolant pumps operating, automatic trip will occur if the main coolant pump motor current is outside the limits on any operating pump. The setpoints specified are consistent with the value assumed in the accident analysis.

Main Coolant System Low Pressure The Main Coolant System Low Pressure trip is provided to prevent operation in the pressure range in which DNBR is less than 1.30 ensuring that the thermal and hydraulic limits assumed in the accident analysis are This Low Pressure trip provides protection by tripping the l

not exceeded.

reactor in the event of a loss of main coolant pressure.

Pressurizer High Water Level The Pressurizer High Water Level trip ensures protection against Main Coolant System overpressurization by limiting the water level to a volume sufficient to retain a steam bubble, prevents water relief through the pressurizer safety. valves, and provides core protection for an uncontrolled rod withdrawai incident or loss of load accident.

B 2-4 YANKEE-ROWE

LIMITING SAFETY SYSTEM SEITINGS BASES Steam Generator Water Level The Low Steam Generator Water Level trip provides core protection by preventing operation with the steam generator water level below the minimum volume required for adequate heat removal capacity. The specified setpoint provides allowance that there will be suf ficient water inventory in the steam generators at the time of trip to provide 15 minutes, as assumed in the accident analysis, for starting delays of the emergency feedwater system.

Turbine and Generator Trip A Turbine or Generator Trip causes a direct reactor trip when operating above 15 MWE. Each of the turbine trips provide turbine protection and reduce the severity of the ensuing transient. No credit was taken in the accident analyses for operation of these trips. Their functional capability is required to enhance the overall reliability of the Reactor Protection System.

Main Steam Isolation Trip A Main Steam Isolation Trip closes the main steam line non-return valves and causes a direct reactor trip. This trip reduces the severity of the cooldown and the ensuing transient effects resulting from a main steam line break. Its functional capability enhances the overall reliability of the Reactor Protection System.

Main Coolant System High Pressure The Main Coolant System High Pressure trip is provided to ensure protection against main coolant system overpressurization caused by a loss of load incident. Its functional capability enhances the overall reliability of the Reactor Protection System.

YANKEE-ROWE B 2-5

3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTDOWN MARCIN LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be > 5.5% AK/K, for Main Coolant Core

~

Average Temperatures 2;515 F.

The SHUTDOWN MARGIN shall be > 4.72% AK/K, for Main Coolant Core Average Temperatures < 485*F.

The SHUTDOWN MARGIN requirement is a linear function between 485 F and 515 F.

APPLICABILITY: MODES 1, 2*, and 3.

ACTION:

With the SHUTDOWN MARGIN less than required, immediately initiate and l

continue boration at ]t 26 gpm of 2200 ppm boron concentration or equivalent until the required SHUTDOWN MARGIN is restored.

SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be 2;that required:

l Within one hour after detection of an inoperable control rod (s) a.

and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter while the rod (s) is l

inoperable. If the inoperable control rod (s) is immovable or untrippable, the above required SHUTDOWN MARGIN shall be increased by an amount at least equal to the withdrawn worth of the immovable or untrippable control rod (s).

b.

When in MODES 1 or 2, at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by verifying that control bank withdrawal is within the limits of Specification 3.1.3.5.

c.

When in MODE 2##, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to achieving reactor criticality, by verifying that the predicted critical control rod position is within the limits of Specification 3.1.3.5.

~ d.- Prior to initial operation above 5% RATED THERMAL FO'w'ER after each fuel loading, by consideration of the factors of e below, with the control banks at the maximum insertion limit of Specification 3.1.3.5.

• See Special Test Exception 3.10.1 i With K,gg 2;1.0 ffWith K,gg < 1.0 YANKEE-ROWE 3/4 1-1

_-- i

._ -.= 2..._: =2 -------'. ~ -

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OPERATION -

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0 10 3D

$0 7d 9'O CONTROL ROD GROUP C POSITION (INCHES WITHDRAWN)

• Allowable THERMAL Power based on the main doolant pump combination in operation.

FIGURE 3.1-1 YANKEE-ROWE 3/4 1-29 e

YANKEEE R0WE RLLOWRBLE PERK R00 LHGR VERSUS CYCLE BURNUP g

12 0 l

i ni==i-i i

i i

i i

i m

m.

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s-w s

r 5 11 0-_

FRESH FUEL e

o r

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w s

r a

taJ g

a.

g HIGH POWER EXPOSED FUEL

_; 10.0; m

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9.5;L

.........$...................d........,...............,,i,,,,.....:i,........-

9.0-e i

(

E, 1) th 14 13 l

0 CYCLE RVERAGE BURNUP (1000 MWD /MTU)

ALL7!fBLE DEAK P09 LHGR VE'1Sl'S CYCLE BUPrUP FIGURE 3.2-1 i

l

.,g-1 1

10--

e

.o u

G w

I L

1 05--

1.00 33 43 53 63 73 83 93 GROUP C POSITION flNCHES WITHDRAWN)

F = F 9 Limit 7

F 0 Measurement FIGURE 3.2-2 Factor F as a Function of Rod Insertion YANKEE ROWE 3/4 2-5 t

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TABLE 3.3-1 REACTOR PROTECTIVE SYSTEM INSTRUMENTATION m

g MINIMUM TOTAL NO.

CHANNELS CHANNELS APPLICABLE m

FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE HODES ACTION 1.

Manual Reactor Trip 3

1 3

1, 2 and

• 1 2.

Power Range, Neutron Flux and Intermediate Power Range, 1, 2 and *(1) 2**

Neutron Flux 6

2 4

3.

Intermediate Range, Neutron Flux, 1(2), 2 and

• 3 High Startup Rate 2

1 2

4..

Source Range, Neutron Flux 2

NA 2

2 and *( )

4 Startup##

a.

ta b

b.

Shutdown 2

NA 1

3, 4, 5(5) 5 5.

Low Main Coolant Flow (SG P) 4 2

3 1(3) 6**

6.

Low Main Coolant Flow (MC Pump Current) a.

System A 4

2 3

1(3) 7**

b.

System B 4

2 3

1(3) 7, CO) 6**

7.

High Main Coolant System Pressure 3

2 3

1, 2 8.

Low Main Coolant System Pressure 3

2 3

1, 2(4) 6**

1, 2( )

8 i

9.

High Prescurizer Water Level 1

1 1

10.

Low Steam Generator Water Level 4

2 3

1(3) 6**

TABLE 3.3-1 (Continued)

REACTOR PROTECTIVE SYSTEM INSTRUMENTATION e

MINIMUM l

TOTAL NO.

CHANNELS CHANNELS APPLICABLE M

FUNCTIONAL UNIT OF CHANNELS TO TRID OPERABLE MODES ACTION 11.

Turbine Trip 1

1 -

1 1(3)(0) 8

12. Generator Trip 1

1 1

1(3)(7) 8

13.. Reactor Trip Breaker 2

1 2

1, 2 and

• 9 14.

Automatic Trip Ingic 2

1 2

1, 2 and

• 9 1, 2(4) 6**

R

15. Main Steam Isolation Trip togic 2

1 2

w a

[

d TABLE 4.3-1 REACTOR PROTECTIVE SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS mE CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED 1.

Manual Reactor Trip NA NA S/U(I)

NA 2.

Power Range, Neutron Flux and Intermediate Power Range, D(2), Q(5)

M 1, 2 and

• Neutron Flux S

3.

Intermediate Range, Neutron Flux, High Startup Rate S

R(5)

M 1, 2 and *

{

4.-

Source Range, Neutron Flux S

R(5)

S/U(I) 2, 3, 4, 5 and

• 5.

Low Main Coolant Flow (SCAP)

S R(0)

M(3) 1 4

6.

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

S R

M 1

7.

High Main Coolant System Pressure S

R(0)

M 1, 2 8.

Low Main Coolant System Pressure S

RIO)

M 1, 2 1

9.

High Pressurizer Water Level S

R(4) g(3) 1, 2 10.

Low Steam Generator Water Level S

RIO)

M 1

~.

TABLE 4.3-1 (continued)

REACTOR PROTECTIVE SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS

. g CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE to CHECK CALIBRATION TEST REQUIRED FUNCTIONAL UNIT NA NA S/U(1) 1

11. Turbine Trip S/U(1) 1 NA NA 12.

Generator Trip

13. Reactor Trip Breaker NA NA S/U(1) 1, 2 and *

14.. Automatic Trip logic NA NA S/U(1) 1, 2 and *

15. Main' Steam Isolation Trip Logic NA NA R

1, 2 Y.

4 e

TABLE 3.3-2 ENGINEERED SAP 7 GUARDS SYSTEM INSTRUMENTATION d

MINIMUM TOTAL NO.

CHANNELS CHANNELS m

.),

OF CHANNELS AND SENSORS AND SENSORS APPLICABLE FUNCTIONAL UNIT AND SENSORS TO TRIP OPERABLE MODES ACTION 1.

SAFETY INJECTION Actuation Channel #1 1

1 1

1, 2, 3(2)(3) 10 a.

1) RPS. Low Main Coolant Pressure Channel 1

1 1

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

2) High Containment Pressure Sensor 1

1 1

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

3) Manual Initiation i

1 1

b.

Actuation Channel #2 1

1 1

1, 2, 3(2) 10 w3

1) Low Main Coolant Pressure Sensor 1

1 1

1, 2, 3(2) 10 w

b

2) High Containment Pressure Sensor 1

1 1

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

3) Manual Initiation 1

1 1

2.

CONTAINMENT ISOLATION a.

Manual Initiation 1

1 1

1, 2, 3, 4, 5(I) 10 CI) b.

Actuation Channel 1

1 1

1,2,3,4,S 10

1) High Containment 1, 2, 3, 4, 5(1).

10 Pressure Sensor 2

1 2

3.

MAIN STEAM IS01ATION a.

Low Steam Line Pressure 3/ Steam Line 2/ Steam Line 3/ Steam Line 1, 2 6**

b.

Automatic Trip logic 2

1 2

1, 2(4) 6**

c.

Manual Initiation 2

1 2

1, 2 6**

d.

High Containment Pressure Trip Containment Isolation 2

1 2

1, 2 6**

TABLE 3.3-2 (continued)

TABLE NOTATION The provisions of Specification 3.0.4 are not applicable.

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

< 300 psig.

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

< 1800 psig.

(3) Automatic initiation of Actuation Channel #1 may be bypassed in this MODE during functional test of the Main Coolant System pressure channel.

I') Trip may be manually bypassed when the reactor is not critical.

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 <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />; however, one safety injection channel high containment pressure sensor may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing per Specification 4.3.2.1.

ACTION 6 - With the number of OPERABLE channels one less than the Total Number of Channels, STl.RTUP and POWER OPERATION may proceed provided both of the following conditions are satisfied:

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

2. The Minimum Channels OPERABLE requirement is met; however, one additional channel may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> I

for surveillance testing per Specification 4.3.1.1.

l l

i YANKEE-ROWE 3/4 3-13

TABLE 3.3-3' g-ENCINEERED SAFECUARDS SYSTEM INSTRUNENTATION TRIP SETPOINTS

~

FUNCTIONAL UNIT-TRIP SETPOINT ll 1.

SAFETY INJECTION a.

Actuation Channel #1

1) RPS Low Main Coolant Pressure Channel 2;1700 psig

2) High Containment Pressure Sensor j[5 psig

3) Manual Initiation Not Applicable b.

Actuation Channel #2

1) Low Mr solant Pressure Sene

> 1700 psig 2)

L..

,ontainment Pressure us Sensor

< 5 psig

,n u,

3) Manual Initiation Not Applicable O.

2.

CONTAINMENT ISOLATION a.

Manual Initiation Not Applicable b.

Actuation Channel

1) High Containment Pressure Sensor j[5 psig 3.

MAIN STEAM ISOLATION a.

Low Steam Line Pressure 2;200 psig b.

Automatic Trip Logic Not Applicable c.

Manual Initiation Not Applicable d.

High Containment Pressure Trip-Containment Isolation 3,5 psig

TABLE 4.3-2' g

ENGINEERED SAFECUARDS SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL MODES IN WHICH

{

CHANNEL CHANNEL FUNCTIONAL

.iURVEILLANCE rd FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED 1.

SAFETT INJECTION a.

Actuation Channel #1 S

NA M(1) 1, 2, 38

1) RPS Low Main Coolant 1

Pressure Channel S

R(3)

M(2) 1, 2, 3#

2). High Containment Pressure Sensor S

R(3)

M(3) 1, 2, 3f

3) Manual Initiation rl. A.

N.A.

R 1, 2, 3, 4, 5*

b.

Actuation Channel #2 S

N.A.

M(1) 1, 2, 3f M

1) Low Main Coolant-Pressure Sensor S

R(3)

M(3) 1, 2, 3f

[

2) High Containment Pressure Sensor S

R(3)

M(3) 1, 2, 37

3) Manual Initiation N.A.

N.A.

R 1, ?, 3, 4, 5*

2.

CONTAINMENT ISOIATION a.

Manual Initiation N.A.

N.A.

R 1, 2, 3, 4, 5*

b.

Actuation. Channel S

N.A.

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

1) High Containment Pressure Sensor S

R(3)

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

3.

MAIN STEAM IS0!ATION a.

Low Steam Line Pressure S

R(3)

M(3) 1, 2 b.

Automatic Trip logic N.A.

N.A.

R 1, 2 c.

Manual Initiation N A.

N.A.

R 1, 2 d.

High Containment Pressure Trip N.A.

E'. A.

R 1, 2

TABLE 3.6-1 (Continued)

CONTAINMENT ISOLATION VALVES TESTABLE DURING VALVE NUMBER FUNCTION PLANT OPERATION ISOLATION TIME (Yes or No)

Second s A.

AUTOMATIC ISOLATION VALVE (Continued)

TV-406

• Main Steam Drain to Condenser No 30 TV-411*

Atmospheric Steam Dump Yes 30 B.

CHECK VALVES SI-V-14*

Safety Injection (HP)

NA NA CS-V-621*

Safety Injection (LP)

NA NA d

T CH-V-611*

MC Feed to Loop #4 NA NA U

CC-V-667*

Component Cooling to MCP #1 NA NA CC-V-663*

Component Cooling to MCP #2 NA NA CC-V-671*

Component Cooling to MCP #3 NA NA CC-V-675*

Component Cooling to MCP #4 NA NA CC-V-649*

Component Cooling to Sample Cooler NA NA CC-V-653*

Component Cooling to Neutron Shield Tank Coolers NA NA CC-V-660*

Neutron Shield Tank Fill NA NA

• Not subject to Type C tests

i 4

TABLE 3.6-1 (Continued) h M

CONTAINMENT ISOLATION VALVES TESTABLE DURING 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-7 54

• Neutron Shield Tank-Inner Test NA NA BF-V-4-1 Air Purge Inlet NA NA yy BF-V-4-2 Air Purge Outlet NA NA

[.

HC-V-602 Air Purge Bypass NA NA s.

SI-MOV-516 ECCS Recirculation No NA SI-MOV-517 ECCS Recirculation No NA BF-CV-1000

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

SC#2 Feedwater Regulator No 30 BF-CV-1200

• SG#3 Feedwater Regulator No 30 BF-CV-1300*

SG#4 Feedwater Regulator No 30 NRV-405A*

Main Steam Non-Return Valve No 5

NRV-405B*

Main Steam Non-Return Valve No 5

NRV-405C*

Main Steam Non-Return Valve No 5

NRV-405D*

Main Steam Non-Return Valve No 5

• Not subject to Type C tests.

J TABLE 3.6-1 (Continued) 4 CONTAINMENT ISOLATION VALVES TESTABLE DURING PLANT OPERATION ISOLATION TIME FUNCTION N

VALVE NUMBER (Yes or No)

Seconds C.

MANUAL VALVES (Cont'd)

NA I/A Main Coolant Heise Pressure Gauge PR-V-610 PU-V-543 Purification System Containment li NA Sump Suction PU-V-544 Purification System Containment N, a NA Sump Suction NA NA EBF-MOV-557 Alternate S.G. Feed w

NA NA N

MS-V-627 * *

• Main Steam Bypass NA NA MS-V-628***

Main Steam Bypass NA NA 4

MS-V-629***

Main Steam Bypass NA NA g

Main Steam Bypass MS-V-630***

NA NA AS-V-719' Emergency Feed Pump Steam Supply NA NA AS-V-720*

Steam Drain 4

Not subject to Type C tests.

• • • Valve may be open for a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> period during secondary plant heat-up and pressure equalization in Modes 2 and 3.

Not subject to type C tests.

I

PLANT SYSTEMS i

EMERGENCY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 At least two independent emergency feedwater pumps and associated flow paths shall be OPERABLE with:

One feedwater pump capable of being powered from an emergency a.

bus, and b.

One feedwater pump capable of being powered from an OPERABLE steam supply system.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

With one emergency feedwater pump inoperable, restore at least two emergency feedwater pumps (one capable of being powered from an emergency bus, and one capable of being powered by an OPERABLE steam supply system) to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

YANKEE-ROWE 3/4 7-5

PLANT SYSTEMS MAIN STEAM NON-RETURN VALVES LD11 TING CONDITION FOR OPERATION 3.7.1.5 Each main steam non-return valve shall be OPERABLE.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

MODES 1 - With one main steam non-return valve inoperable, POWER OPERATION may continue provided the inoperable valve is either restored to OPERABLE status or closed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

MODES 2 - With one main steam non-return valve inoperable, subsequent l

and 3 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 hourc. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.1.5 Each main steam non-return valve that is open shall be demonstrated l

OPERABLE by:

Cycling each valve through at least 10% of full travel at least a..

l once. per 92 days, and b.

Verifying full closure within 5 seconds on any closure actuation l

signal whenever shutdown longer than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, if not performed in the previous 92 days.

YANKEE ROWE 3/4 7-9

3/4.1 REACTIVITY CONTROL SYSTEMS nASES 3/4.1.1 BORATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made suberitical from all operating conditions, 2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and 3) the reactor will be maintained sufficiently suberitical to preclude inadvertent criticality in the shutdown condition.

l SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, Main Coolant System boron concentration, and Main Coolant The most restrictive condition occurs at EOL, with T at SystemT,y$r.ating temperature, and is associated with a postulated,afeam no load op line break accident and resulting uncontrolled Main Coolant System cooldown.

In the analysis of this accident, a minimum SHUTDOWN MARGIN of 4.72% dk/k is initially required to control the reactivity transient. Accordingly, the SHUTDOWN MARGIN requirement is based upon this limiting condition and is consistent with accident analysis assumptions. The value of 5.5% ak/k is incorporated to provide added SHUTDOWN MARGIN and reflects the actual excess of shutdown margin availabic at the plant. With T yg i 330 F, the reactivity transients resulting from a postulated steam 1 ne break cooldown are minimal. 5% Ak/k SHUTDOWN MARGIN (with all rods inserted) provides adequate protection to preclude criticality for all postulated accidents with the reactor vessel head in place.

j To eliminate possible errors in the calculations of the initial reactivity of the core and the reactivity depletion rate, the predicted relation between fuel burnup and the boron concentration, necessary to maintain adequate control characteristics, must be adjusted (normalized) to accurately reflect actual core conditionc. Normally, when full power is reached after each refueling, and with the control rod groups in the desired positions, the boron concentration is measured and the predicted steady state curve is adjusted to this point. As power operation proceeds, the measured boron concentration is compared with the predicted concentration and the slope of the curve relating burnup and reactivity is compared with l

that predicted. This process of normalization should be completed after about 10% of the total core burnup. Thereafter, actual boron concentration can be compared with prediction and the' reactivity status of the core can be continuously evaluated ' and any deviation would be thoroughly investigated and evaluated.

YANKEE-ROWE B 3/4 1-1 L

PLANT SYSTEMS BASES 3 /4. 7.1.4 ACTIVITY The limitations on secondary system specific activity ensure that the resultant off-site radiation dose will be limited to a small fraction of 10 CFR Part 100 limits in the event of a steam line rupture. This dose also includes the effects of a coincident 1.0 GPM primary to secondary tube leak in the steam generator of the affected steam line. These values are consistent with the assumptions used in the accident analyses.

The steam break accident is based upon a postulated release of the entire contents of the secondary system to the atmosphere using a site boundary dose limit of 1.31 rem for thyroid dose.

The limiting dose for this accident results from iodine in the secondary coolant. The reactor distribution of iodine isotopes with 1%

failed fuel was used for this calculation. 1-131 is the dominant isotope because of its low MPC in air and because the other iodine isotopes have shorter half-lives and therefore cannot build up to significant concentrations in the secondary coolant, given the limitations on primary systemleakrateagdactivity. The entire secondary system contains approximately 132m of water at standard conditions. One-tenth of the contained iodine is assumed to reach the site boundary, making allowance for plate-out and retention in water droplets.

3/4.7.1.5 MAIN STEAM NON-RETURN VALVES The OPERABILITY of the main steam non-return valves ensures that no more than one steam generator will blowdown in the event of a steam line rupture. Main steam non-return valve auto-closure minimizes the Main Coolant System cooldown associated with the blowdown.- This feature enhances plant performance by:

1) Minimizing the reactivity transient.

2) Minimizing the Main Coolant and Secondary System thermal transient.

3) Providing additional backup to normal non-return action as a check valve to limit the containment transient resulting from a main steam line rupture inside the containment.

YANKEE-ROWE B 3/4 7-3

5.0 DESIGN FEATURES 5.1 SITE EXCLUSION AREA 5.1.1 The exclusion area shall be as shown in Figure 5.1-1.

LOW POPULATION ZONE 5.1.2 The low population zone shall be as shown in Figure 5.1-2.

5.2 CONTAINMENT CONFIGURATION The reactor containment building is a steel spherical shell having 5.2.1 the following design features:

Nominal inside diameter = 125 feet.

a.

b.

Minimum thickness of steel shell = 7/8 inches.

Net free volume = C00,000 cubic feet.

c.

DESIGN PRESSURE AND TEMPERATURE The reactor containment is designed and shall be maintained for a 5.2.2 0

maximum internal pressure of 34.5 psig and a temperature of 249 F.

5.3 REACTOR CORE FUEL ASSEMBLIES The reactor core shall contain 76 fuel assemblies with each fuel 5.3.1 Each fuel assembly containing 230 or 231 fuel rods clad with Zircaloy-4.-

Each f.nel assembly rod shall have a nominal active fuel length of 91 inches.

Reload fuel shall contain a may %um total weight of 234 kilograms uranium.

is similar in physical design to the Core XII EXXON fuel and shall have a maximum enrichment of 3.5 weight percent U-235.

4 5-1 YANKEE-ROWE

~

United Statao Nuclear Regulatory Commission March 26, 1981 At ten tion: Office of Nuclear R2setor Regulation Page 3 ATTACHMENT B NON-RETURN VALVE MODIFICATION P

The four (4) main steam line non-return valves (NRV) will be modified to provide automatic, quick closure of the NRV's.

This will be accomplished by implementation of a quick closing (3-5 seconds) stored energy actuator on each o f the f our main s team NRV's. These actuators will be provided by Rockwell International with conversion kits to permit installation on the existing In addition, this modification will also necessitate (1) installation I

valv es.

(2) of a new steam supply line to the steam driven emergency feedwater pump, installation of additional atmospheric steam dump capacity, and (3) installation of a weatherproof enclosure to assure a controlled temperature envir onment.

The non-return valves will automatically isolate on low steam line pressure or a containment isolation signal on high containment pressure.

Each steam line will be fitted with three pressure switches upstream of the NR'.* in th a t line.

On low steam line pressure, each of the three pressure switches will operate a set of two relays which will be powered by three independent 125V de batteries. Thus, there will be three channels associated with each steam line. Contacts from the channel relays will be combined to The produce a 2 out of 3 (2/3) trip logic to generate the NRV trip signal.

2/3 trip logic will also generate a permissive signal to trip the condensate Channel relays will be energized to actuate a trip.

pumps.

The containment isolation system will provide a high containment pressure signal to generate the NRV trip signale This trip signal will operate in the presence of either the steam line pressure signal, or the high containnent pressure signal.

The NRV trip logic will be redundant, that is, there will be a Train A and a Train B trip circuit. Each train will be powered by an independent 125V de The two channel relays associated with each pressure switch above battery.

will be used such that the contacts of one -elay will be for the Train A trip circuit and contacts of the other will be for the Train B trip circuit.

A high containment pressure signal will be provided from a relay in each Containment Isolation System _(CIS) train. The Train A CIS relay contact will produce a trip in the Train A NRV trip circuit and the Train B CIS relay The NRV trip contact will produce a trip in the Train B NRV trip circuit.

circuit relays will be energized to cause a trip.

The Technical Specification changes are necessary due to the addition of the main' steam line isolation circuitry and instrumentation as well as to incocporate revised surveillance requirements.

This modification reduces the severity of transients imposed on the main coolant system and reactor vessel in the event of the main steam line break design basis event (DBE). Quick auto closure of the NRV's would ensure the availability of an adequate heat sink in the event of the DBE.

The reliability of the stored energy actuator is high,- since the closing

United Statao Nuclear Regulctory Commission March 26, 1981 At ten tion: Offica of Nuclser Reactor Regulation Page 4 energy for the valve is self-contained.

No external safety-related power source. other than that required for ins trumenta tion, is necessary to perform the closing function. All safety-related systems internal and external to the actuator are designed for single active failure.

This change does not present a significant hazard to the public health and safety, nor does it constitute an unreviewed safety question as described in 10 CFR 50.59 (a) (2).

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United States Muclear Regulatory Commiesion March 26, 1981 Attsnticn: Office of Nuclear Reactor R3gulation Page 5 ATTACHMENT C EMERGENCY FEEDWATER SYSTEM MODIFICATION This modification increases the existing emergency feedwater capab'ilities by adding two full capacity motor driven pumps. Additional piping will be installed to allow feeding the steam generators with either pump through the normal feedwater piping via the steam driven emergency feedwater pump discharge header, or through the blowdown piping via the alternate emergency feedwater header.

System actuation will be capable of being initiated from the control room or locally by operator action.

The two motor driven pumps will be located in the Primary Auxiliary Building and will have the capability to take suction from either the primary water storage tank, TK-39, or the demineralized water storage tank, TK-1.

The pump suctions will be normally aligned to TK-39.

The new pumps will use existing 300 HP motors. These motors will be powered f rom the 2400 V busses No. 2 and 3.

Remote flow indication to each steam generator will be available in the control room for emergency feedwater flow through the normal f eedwater path Local flow indication will be available in the combined pump minimum recirculatien flow piping.

Technical Specificati,on changes are necessary to incorporate the new pumps, revise the surveillance requirements, and revise the containment isolation valve list due to the piping modification.

This design change increases the capability and the redundancy of the emergency feedwater system, resulting in an increased system reliability. The revised system will consist of three pumps, two motor driven and one steam turbine driven, capable of feeding the steam generators through two separate feed paths. Also, the capability will now exist to initiate emergency feedwater, via the two motor driven pumps, manually from the control room.

This change does not present a significant hazard to the public health and safety nor does it constitute an unreviewed safety question as described in 10 CFR 50.59(a)(2).

O ATTACHMENT D CORE XV PERFORMANCE ANALYSIS REPORT L