Text

Amend 69 to License DPR-3,incorporating Provisions Into Tech Specs Required for Operation W/Refueled Core Xv & Associated W/Mods to Main Steam Nonreturn Valves & Auxiliary Feedwater Sys
	ML20009F329
Person / Time
Site:	Yankee Rowe
Issue date:	07/22/1981
From:	Crutchfield D; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20009F328	List: ML20009F327; ML20009F329; ML20009F330; ML20009F332;
References
	NUDOCS 8107300299
	Download: ML20009F329 (57)
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UNITED STATES gs y '

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NUCLEAR REGULATORY COMMISSION g-Jj WASHINGTON. D. C. 20555

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THE YANKEE ATOMIC ELECTRIC COMPANY i

DOCKET Nn. 50-29 YANKEE NUCLEAR POWER STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 69 License No. DPR-3 1.

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

A.

The application for amendment by the Yankee Atomic Electric Company (the licensee) dated March 26, 1981, as supplemented May 27,1981 and July 8,1981, 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 CFh 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 (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance witt ':he Commission's regulations; D.

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.

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G107300299 810722 PDR ADOCK 05000029 P

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Accord'ingly, the license is amended by changes '.o - the Technical Specifications as indicated in the attachment.to this license amendment,.and Paragraph 2.C(2) of Facility,0perating License No. OPR-3 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in-Appendix A, as revised.through Amendment No. 69, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Techni-

. cal Specifications.

3.

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

FOR THE NUCLEAR EG ATORY COMMISSION

^

d Dennis c field, Chief Operating _ Reactors Branch #5 Division of Licensing

Attachment:

Changes to the Technical Specifications

(

Date of Issuance:

July 22', 1981 i

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.1 ATTACHMENT TO LICENSE AMENDMENT NO. 69 FACILITY OPERATING LICENSE NO. DPR-3 DOCKET NO.'50-29 Replace the following pages.of the Appendix "A" Technical Specifications with the enclosed pages as indicated. The revised pages are identified

- by Amendment Number and contain. vertical lines Indicating the area of

-' change.

DELETE INSERT IV IV*

VI

-VI l-5 1-5**

1-6 1-6 2-1 2-l**

2-2 2-2 2-3 2-3 2-4 2 4**

2-5 2-5**

4 2-6 2-6

'B2-3 B2-3**

B2-4 B2-4 j

B2-5 B2-5 3/4 1-1 3/4 1-1 3/4 1-2 3/4 1-2 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 2-6 3/4 2-6 3/4 2-7.

3/4 2-7 i

3/4 2-8 3/4 2-8**

3/4 3-1 3/4 3-l**

3/43-2 3/4 3-2 3/4 3-3 3/4 3-3 3/43-4 3/4 3-4**

3/a 3-7 3/4 3-7**

l 3/4 3-8 3/4 3-8 3/4 3-9 3/4 3-9 3/4 3-10 3/4 3-10**

3/4 3-12A 3/4 3-12A 3/4 3-13 3/4 3-13 3/4 3-14 3/4 3-14 3/4 3-15A 3/4 3-15A i

This is included to merely correct editorial-errors.

Overleaf page ir.cluded for' completeness of records.

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- INSERT 3/4 5-5' 3/4 5-5 3/4 5-6 3/4 5-6 3/4'6-11 3/4 6-11**-

.3/4 6-12 3/4 6-12 3/4 6-13 3/4 6-13**

3/4 6-14 3/4 6-14

~2/4 6-14a (new) 3/47-5 3/47-5 3/4 7-Sa (new);

3/4 7-9

.3/4 7-9 B3/4 1-1~

B3/4 1-1 B3/4 1-2 B3/4 1-2**

83/4'2-4 B3/4 7-1**

B3/4 7-2 B3/4 7-2 B3/4 7-3 83/4 7-3 5-1 5-1 5-2 5-2**

5-3 5-3**

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Overleaf page included for completeness of records.

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LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS I

SECTION Page 3/4.2 P0'.ER DISTRIBUTION LIMITS 3/4.2.1 PIAK LINE AR HEAT CENE RATION RATE......................... 3 /4 2-1 3/4.2.2 NUCLEAR HEAT FLUX HOT CHANNEL FACTOR..................... 3 /4 2-7 3/4.2.3 NUCLE AR ENTHALPY RI SE HOT CHANNEL FACTOR................. 3/4 2-9 3/4.2.4 DN B P ARAMEIE RS........................................... 3 /4 2-11 3/4.3 INS. HENTATION 3/4.3.1 FI ACTOR PROTECTIVE SYSTEM INSTRUMENTATION................

3/4 3-1 3/4.3.2 EN0!NEEFID SAFECUARDS SYSTEM INSTRUMENTATION............. 3/4 3-11 3/4.3.3 MONITORING INSTRUMENTATION 1

Radiation Monit oring Ins trumentation..................... 3/4 3-17 Incore Detection System.................................. 3/4 3-23 Met eo rolo gic al Ins trume nt a tion........................... 3 /4 3-24 Fire Detection Instrumentation........................... 3/4. 3-27

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Accident Monitoring Instrumentation 3/4 3-29 2

, / ~. 4 MAIN COOLA:." SYSTod 3

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3/4.4.1 MAIN C001 ANT LOOPS j

C o o l an t Ci rc ul a tio n......................................

3 /4 4-1 l

l Isolated Loop............................................ 3/4 4-3 Main Coolant Loop Startup................................

3/4 4-4 3/4.4.2 SAFETY VALVE S - SFUTD0' *N................................. 3 /4 4-3 *

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3/4.4.3 S AFETY VALVE S - OPERATING................................

3 /4 4-6 3/4.4.4 P FI S SU RI L'E R.............................................. 3 / 4 4 - 7 3/4.4.5 GIN COOLANT SYSTEM LEAKAGE j

Lea ka ge Det ec tio n S y s t ems................................

3 /4 4-8 O erational Leakage..................................... 3/4 4-10 P

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With 3/4 4-Sa gy Yankee-Rowe Amendment No. 46, 59, 65, 66, 59

o INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIkFMENTS j

PAGE SECTION 3/4.6.3 COMBUSTIBLE CAS CONTROL i

Hyd roge n Anal y ze r.................................... 3/4 6-16 H yd roge ri Ve n t Sw s t em................................. 3/4 6-17 Atmosphere Recirculation System......................

3/4 6-18 3/4.7 PLANT SYSTEMS 4

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

3/4 7-1 i

Eme rgency Feedwa te r Sys tem........................... 3/4 7-5 Primary and Demineralized Water Storage Tanks........ 3/4 7-6 Activity.............................................

3/4 7-7 i

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

3/4 7-9 l

3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION...... 3/4 7-13 l

3 /4.7.3 PRIMARY PUMP SEAL WATER SYSTEM (Deleted).............

3/4 7-1*

3 /4. 7.4 SERVICE WATER SYSTEM (Dele ted)....................... 3/4 7-16 3 /4.7. 5 CONTROL ROOH VENTILATION SYSTEM EMERGENCY SHUTDOWN... 3/4 7-18 3/5.7.6 SEALED SOURCE CONTAMINATION..........................

3/4 7-19 3 /4. 7. 7 WASTE EFFLUENTS Rad io a c t ive S;11d Wa s t e.............................. 3/4 7-21 Rad ioac t ive Li c ' ' i Wa s t e............................. 3/4 7-22 i

Radioactive Gsseous Waste............................

3/4 7-23 3/4.7.8 ENVIRONMENTAL MONITORING....... <..................... 3/4 7-24 3/4.7.9 SHOCK SUPPRESSORS (SNUBBERS )......................... 3/4 7-27 3/4.7.10 FIRE SUPPRESSION SYSTEMS.............................

3/4 7-30 3/4.7.11 PENETRATION FIRE BARRIERS............................

3/4 7-37 r

VI Amendment No. 47, 52, 56, 51, 64, 69 YANKEE-ROWE

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DEFINITIONS the sum of the average beta and gama energies per disintegration (in MeV) for isotopes, other than iodines, with half lives greater than 30 minutes, making up at least 95% of the total non-iodine activity in the coolant.

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TABLE 1.1 OPERATIONAL MODES REACTIVITY

% RATED AVERAGE *COOLAhT MODE CONDITION K,gg THERMAL POWER *

, TEMPERATURE __

0 1.

POWER'0PERATION 10.99

> 2%

1 330 F 0

2.

STARTUP

> 0.99

< 2%

> 330 F 0

3.

HOT STANDBY

< 0.99 0

1 330 F 0

4.

HOT SHUTDOWN

< 0.96 0

330 F > T,yg > 200 F 0

5.

COLD SHUTDOUN

< 0.96 0

1 200 F 0

6.

REFUELING **

10.95 0

1 140 F d

Excluding decay heat.

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

- OE 1-6 Amendment No. 69

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS i

2.1 _ SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, Main Coolant System pressure, and the highest operating loop cold leg coolant temperature shall not exceed the limits shown in Figures 2.1-1 and 2.1-2 for 4 and 3 loop operation, respectively.

APPLICABILITY: MODES 1 and 2.

ACTION:

Whenever the point defined by the combination of the highest operating loop cold leg temperature and THERMAL POWER has exceeded (is above and to the right of) the appropriate Main Coolant System pressure line, be in HOT STANCBY within 1 nour.

MAIN COOLANT SYSTEM PRESSURE 2.1.2 The Main Coolant System pressure shall not exceed 2735 psig, i

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

ACTION:

MODES 1 and 2 Whenever the Main Coolant Systea pressure has exceeded 2735 psig, be in HOT STANDBY with the Main Coolant System pressure within its limit within I hour.

MODES 3, 4 and 5 Whenever the Main Coolant System pressur'e has exceeded 2735 psig, reduce the Main Coolant System pressure to within its limit within 5 minutes.

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FIGUPI 2.1-2 Amendment No. 43, 59 YANKEE-ROWF.

2-3

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR PROTECTIVE SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The reactor protective system instrumentation setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2-1.

APPLICABILITY: As shown for each channel in Table 3.3-1.

_ ACTION:

With a reactor protective system instrumentation trip setpoint less conservative than the value shown in the Trip Setpoint column of Table 2.2-1, declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1 until the channel is restcred to OPERABLE status with its trip setpoint adjusted consistept with the Trip Setpoint value.

e YANKEE-ROWE 24

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TABLE 2.2-1 2

5%

REACTOR PROTECTIVE SYSTEM INSTRUMENTATION TRIP SETPOINTS 2

T o"

i FUNCTIONAL UNIT TRIP SETPOINT

1. Manual Reactor Trip Not Applicable

2. Power Range, Neutron Flux Low Setpoint - 3,35% of RATED THERMAL POWER High Setpoint - < 108% of RATED THERMAL POWER with 4 main coolant pumps operating High Setpoint - 5,81% of RATED THERMAL POWER with 3 main coolant pumps operating

3. Intermediate Power Range, High Setpoint - 5.108 3f RATED THERMAL POWER with 4 main coolant 7'

Neutron Flux pumps operating m

High Setpoint - 3,81% of RATED THERMAL POWER with 3 main coolant i

pumps operating l

4.

Intermediate Range, High 3,5.2 decades / minute Startup Rate l

S. Source Range, Neutron flux Not Applicable 6.

Low Main Coolant Flow

> 80% of Design Flow (steam generator AP)

)

7.

Low Main Coolant Flow

~'>240 Amperes,<950 Amperes (main coolant pump current) l l

l L______________

TABLE 2.2-1 (continued) r.p REACTOR PROTECTIVE SYSTEM INSTRUMENTATION TRIP SETPOINTS S

b FUNCTIONAL UNIT TRIP SETPOINT 8.

liigh Main Coolant System Pressure j[ 2300 psig -

l-9.

Low Main Coolant System Pressure

}>1800 psig 10.

Illgh Pressurizer Water Level j[200 inches 11.

Low Steam Generator Water level

}> - 13"

12.

Turbine Trip Not Applicable 13.

Generator Trip Not Applicable ya m

14.

Main Steam Isolation Trip Logic

}> 200 psig' N

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

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2.2 LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR PROTECTIVE SYSTEM INSTRUMENTATION SETPOINTS The Reactor Trip Setpoint limits specified in Table 2.2-1 are the values at which the reactor trips are set for each parameter. The Trip Setpoints have been selected to ensure that the reactor core and Main Coolant System are prevented from exceeding their safety limits.

Manual Reactor Trip The Manual Reactor Trip is a redundant channel to the autoratic protective instrumentation channels and provides manual reactor trip capability.

Power Range and Intermediate Power Range, Neutron Flux The Power Range and Intermediate Power Range Neutron Flux channel high setpoint provides reactor core protection against reactivity excursions which are too rapid to be protected by pressurizer water level protective circuitry. The Power Range low set point provides additional protection

/

in the power range for a power excursion beginning from low power. The trip associated with the low setpoint may be manually bypassed above 15 MWe and is manually reinstated at a power level below 15 MWe. The low setpoint trip is not assumed in the accident analysis.

The prescribed setpoint, with allowances for errors, is consistent with the trip point used in the accident analysis. The lower setting for three loop operation provides the protection at the reduced power level equivalent to that provided by the setting for four loop operation at full power.

Intermediate Range, Neutron Flux, High Startup Rate The Intermediate Range High Startup Rate trip provides protection to limit the rate of power increase during low power conditions in the

' event of an uncontrolled rod withdrawal.

YANKEE-ROWE B 2-3 w

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e LIMITING SAFETY SYSTEM SETTINGS BASES Low Main Coolant Flew (Steam Generator dP)

The Low Main Coolant Flow trips prc<1de core protection in the event of a loss of one or more main coolant p2mpc-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 assumed 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 pumps op sting, an outside automatic trip will occur if the main coolant pump motor currea; 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 cperating 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 analysir are not exceeded. This Low Pressure trip provides protection by tripping the l

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

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

YANKEE-ROWE B 2-4 Amendment No. 69

o LIMITING SAFETY SYSTEM SETTINGS 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 sufficient 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 la 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 ef fects resulting from a main steam line break. Its functional capability enhances the overall reliability of the Reactor Protection System.

I 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.

l YANKEE-ROWE B 2-5 Amendment No. 69 n--

I' 3/4.1 REACTIVITY C0!rfROL SYSTEMS l

l 3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION I

3.1.1.1 The SHUTDOWN MARGIN shall be > 5.5% AK/K, for Main Coolant Core

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Average Temperatures 1 515 F.

The SHUTDOWN MARGIN shall be > 4.72% AK/K, for Main Coolant Core Average

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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 > 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 > that required:

l a.

Within one hour after detection of an inoperable control rod (s) and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the rod (s) is inoperable. If the inoperable control rod (s) is immovable or untrippabic, 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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by verifying that control bank withdrawal is within the limits' of Sp6cification 3.1.3.5.

c.

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

d.

Prior to initial operation above 5% RATED THERMAL POWER af ter 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.

l

See Special Test Exception 3.10.1

With K,ff 1 1.0

1. With K,gg < 1.0 Amendment No. 69 YANKEE-ROWE 3/4 1-1

l REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

+

e.

When in MODE 3, at least once per 24 heurs by consideration of the following factors:

1.

Main Coolant System baron concentration, 2.

Control rod position.

3.

Main Coolant System average temperature, 4

Fuel burnup based on gross thermal energy generaticn, 5.

Xenon concentration, and 6.

Samarium concentration.

4.1.1.1.2 The overall core reactivity balance shall be compared to j

predicted values to demonstrate agreement within + 0.8% ek/k at least once per 31 Effective Full Power Days (EFPD). This cdparison shall consider at least those factors stated in Specification 4.1.1.1.1.e, above.

The predicted reactivity values shall be adjusted (normalized) to corre-spond to the actual core conditions prior to exceeding a fuel burnup of r ll Te<.er Days af ter each fuel loading.

60 Effective u

4.1.1.1.3

':nenever the reactor is shut down, before any operation which night result in i change of reactivity, a control rod group shall be withdrawn to a height sufficient, provide a reactivity worth of l'; for e.ergency shutdown capability.

4 f for any reason this is not practical, r

the ,ain Coolant System shall be borated to provide 5% ak/k SHUTDOWN MARGIN with all control rods inserted.

4.i.1.1.4 During a' reactor startup in which core reactivity or control rod positions for criticality are not established, a plot of inverse cultiplication rate (or count rate) versus rod position shall be made.

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FIGURE 3.1-1 YANKEE-ROWE 3/4 1-29 Amendment No. 43, 59 4

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w u

J w

a.

60_

~

s nC

.75_.

E

~

70-vvrr)rrrrJmr]rvrrj....j....j...js ervjrm-jqrr..f....,'....,1....,',....

3 3

g cyctt ExrosuML towo/nrun o

et

=

.O w%

Figure 3.2-4 a"

Multiplier for Reduced Power as a Function of Exposure tn a

b i

--yr

-Wg v--

y www y

a w w-1

-w-r

-tv

+---.- --

y

-,-yv qw-

s POWER DISTRIBUTION LIMITS HEAT FLUX HOT CHANNEL FACTOR-F q

QMITING CONDITION FOR OPERATION shall be limited by the following relationships:

3.2.2 Fq for P > 0.5 q $ [2.76]

F l

F j [5.523 for P y 0.5 q

where P = THERMAL POWERRATED THE MAL T0lTE APPLICABILITY _: MODE 1 ACTI ON_:

With F exceeding its limit:

q exceeds the Reduce THERMAL POWER at least 1% for each 1% F limit within 15 minutes and similiarly reduce Ihe Power Range a.

and Intermediate Power Range Nuetron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , Identify and correct the cause of the out of limit condition prior to increasing THERPAL POWER above the reduced limit b.

required by a, above; THERFAL POWER may then be increased is demonstrated through incere mapping to be provided Fq within its limit.

YANKEE-ROWE 3/4 2 8 A9endment flo.,4, 69

[

~

e.

3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTIVE SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION As a minimum, the reactor protective system instrumentation 3.3.1 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 MODES and at the frequencies shown in Table 4.3-1.

The logic for the Reactor Permissive Circuit shall be demon-4.3.1.2 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 CAllBRATION testing of each channel affected by permissive circuit operation.

YANKEE-ROWE 3/4 3-1

"o 4

TABLE 3.3-1 REACTOR PROTECTIVE SYSTEM INSTRUMENTATION g

S MINIMUM Q

M TOTAL NO.

CHANNELS CHANNELS APPLICABLE-FUNCTIONAL UNIT OF CllANNELS TO TRIP OPERABLE MODES ACTION 1.

Manual Reactor Trip 3

1 3

1, 2 and

1 2.

Power Range, Neutron Flux and Intermediate Power Range, 1, 2 and *II) 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 Startup##

2 NA 2

2 and *I }

4 a.

I5) b.

Shutdown 2

NA 1

3, 4, S 5

1(3)'

6**

5.

Low Main Coolant Flow (SG P) 4 2

3 6.

Low Main Coolant Flow (MC Pump Current) 4 2

3 1(3) 7**

a.

System A b.

System B 4

2 3

1(3) 7,,

7.

High Main Coolant System Pressure 3

2 3

1, 2(4) 6**

8.

Low Main Coolant System Pressure 3

2 3

1, 2(4) 6**

5 I0) 9.

High Pressurizer Water Level 1

1 1

1, 2 8

m 10.

Low Steam Generator Water Level 4

2 3

1(3) 6**

TABLE 3.3-1 (Continued)

REACTOR PROTECTIVE SYSTEM INSTRUMENTATION M

hl MINIMUM M

TOTAL NO.

CllANNELS CilANNELS APPLICABLE FUNCTIONAL UNI' OF CllANNELS TO TRIP OPERABLE MODES ACTION 11.

Turbine Trip 1

1.

I 1(3)(6) 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 Logic-2 1

2 1, 2 and

9 ki 15.

Main Steam Isolation Trip Logic 2

1 2

1, 2( }

6**

z.

m ai

. o 5I

TABLE 3.3-1 (Continued)

TABLE NOTATION

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

- The provisons of Specification 3.0.4 are not applicable.

High voltage to detector is automatically de-energized above 5 x 10-9 Amperes on the Intemediate Range.

1. r when other activities might increese reactivity.

0 (1) Power Range, Neutron Flux, Low Setpoint Trip may be manually bypassed at > 15 MWe. Bypass shall be manually removed at i 15 NWe.

(2)

Intemediate Range, Neutron Flux, High Startup Rate Trip is auto-matically bypassed > 15 MWe.

1 5 MWe.

Bypass is automatically, removed at 1

(3) Trip may be manually bypassed 1 15 MWe.

Bypass is automatically removed at > 15 MWe.

(4) Trip may be manually bypassed when the reactor is not critical.

(5) Startup rate alam setpoint 1 1.1 decade / minute.

(6) Turbine shall be protected by at least the following protective trips:

rotor excessive axial movement, low bearing oil pressure; low condenser vacuum; and overspeed.

(7) Generator shall be protected by at least the following protective trips: overcurrent; differential; and loss of field.

ACTION STATEMENTS ACTION 1 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least h'OT STANDBY within the next 6 hou s and/or open the reactor trip breakers.

YANKEE-ROWE 3/4 3-4

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 thar 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.

YANKEE-ROWE 3/4 3-7 j

O T

TABLE 4.3-1 REACTOR PROTECTIVE SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS A

CHANNEL MODES IN WHICH Q

"8 CilANNEL CllANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CllECK CALIBRATION TEST REQUIRED 1.

Manual Reactor Trip NA NA S/U(l)

M 2.

Power Range, Neutron Flux and Intermediate Power Ratige, 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/UII) 2, 3, 4, 5 and *

{

5.

Low Main Coolant Flow (SCAP)

S R(4)

M(3) y 4

6.

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

S R

M 1

R ')

M 1, 2 I

7.

High Main Coolant System Pressure S

N R(4)

M 1, 2 8.

Low Main Coolant System Pressure S

9.

High Pressurizer Water Level S

R(4)

M(3)

' 1, 2 10.

Low Steam Generator Water Level S

R(4)

M 1

s

l t

TABLE 4.3-1 (continued)

REACTOR PROTECTIVE SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS g

CHANNEL MODES IN WHICH CIIANNEL CHANNEL FUNCTIONAL SURVEILLANCE m

I A

FUNCTIONAL UNIT CilECK CALIBRATION TEST REQUIRED R

S/U(I) 1 11.

Turbine Trip NA NA II) 12.

Generator Trip NA NA S/U 1

13.

Reactor Trip Breaker NA NA S/U(I) 1, 2 and

14.

Automatic Trip Ingic NA NA S/U(I) 1, 2 and

0 1, 2 h

15.

Main Steam Isolation Trip logic NA NA Y

e F

l 2

a i

O l

l

TABLE 4.3-1 (Continued)

NOTATION

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

,l i

If not performed in the previous 7 days.

(1)

Heat balance only, above 15% of RATED THERMAL POWER, at least (2) 3 times per week with a maximem 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.

Known pressure applied to sensor.

(4)

Neutron detectors may be excluded from CHANNEL CALIBRATION.

l (5)

YANKEE-ROWE 3/4 3-10 Amendment No. 49, 69 E

?!

TABLE 3.3-2 (Continued)

El ENGINEERING SAFEGUARDS SYSTEM INSTRUMENTATION E$

ni HINIMUM TOTA!. NO.

CIIANNELS CilANNEI,S OF CilArit1El.S At4D SENSORS.

AND SENSORS APPLICABLE FUt4CTIONAL titilT AtlD set 4 SORS TO TRIP OPERABLE MODES ACTION 2.

C0tiTAllCIEliT ISOLATI0ti (Continued)

c. Actuation Channel B 1

1 1

1,2,3,4, SCI) 10 l) llinh Containment Pressure Sensor I

1 1,2,3,4,Sil) 10

~

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

ff a

y 3.

MAIN STEAM ISOLATION 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.

liigh Containment Pressure Trip s

Containment Isolation 2

1 2

1, 2 6**

g-8 1

n E?

~

8; j

n

)

i

TABLE 3.3-2 (continued)

TABLE NOTATION

- The provisions of Specification 3.0.4 are nnt 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)Autc;atic initiation of Actuation Channel #1 'may be bypassed in this MODE during functional test of the Main Coolant System pressure channel.

(4) Trip may be manually bypassed when the reactor is not critical.

ACTION STATEMENTS ACTION 10 - With the number of OPERABLE channe'.s 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 SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ; however, one safety injection channel high containment pressure sensor may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.2.1.

I l

ACTION 6 - With the number of OPERABT E channels one less than the Total l

Number of Channels, STARTUP and POWER OPERATION may proceed j

provided both of the following conditions are satisfied:

1. The inoperable channel is placed in the tripped condition.

within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2. The Minimun Channels OPERABLE requirement is met; however, l

one additional channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months l

for surveillance tescing per Specification 4.3.1.1.

l l

YANKEE-ROWE 3/4 3-13 Amendment No. 69

TABLE 3.3-3 ENSINEEPID SAFECUARDS SYSTEM INSTRL ENTATION 'DtIP SETPOINTS ITIP SETPOINT FUNCTIONAL UNIT 1.

S A.WIY I?UECTION a.

Actuation Channel #1

1) RPS Low Main Coolant

.,1700 psig Pressure Channel

2) High Containment Pressure Sensor

-j[ 5 psig

3) Manual Initiation Not Applicable b.- Actuation Channel #2

1) Low Main Coolant Pressure Sens or

> 1700 psig

2) High Containment Pressure Se ns or j[ 5 psig

3) Manual Initiation Not Applicable e

2.

CONTAIM'2NT ISOLATION a.

Manual Initiation Not Applicable b.

Actuation Channel A

~

1) High Containment Pressure Se ns or j[5 psig

2) Safety Injection.

(All Safety s'

Injection Setpoints) c.

Actuation Channel B

1) High Contain=ent Pressure j[ 5 psig S ens or ~

2) Safety Injection (All Safety Injection Setpoints) 3.

MAIN STEAM ISOLATION

> 200 psig a.

Low Steam Line Pressure Not Applicable L.

Automatic Trip Logic Not Applicable c.

Manual Initiation d.

High Containment Pressure Trip-Containment Isolation j[5 psig yj_w.II-R3tt!

3/4 3-14 Amendment No.,5(.5o 69-t i

.-~

i TABLE 4.1-2 (Continued) 5 z

ENGINEERED SAFEGUARDS' SYSTEM INSTRUMENTATION h[

SURVEILLANCE REQUIREMENTS A

E m

. Cil ANiiEL HODES IN WilICll CilAtillEL CllAriflEL FUtiCTIONAL SURVEILLANCE FUtiCTint1AL UtilT C11ECI; CALI nitATION TEST REQUIRED 2.

CollTA1HMEtiT ISOLATION (Continued)

c. Actuation Channel n S

ti. A.

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

R.

.1) liigh Containment S

R(3)

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

Pressure. Sensor, Y'

3;.

2) Safety Injection' (All Safety Injection Surveillance Requirements) 3.

MAIN STEAM ISOLATION a.

Low Steam Line Pressure S

R(3)

.M(3) 1, 2 b.

Automatic' Trip Logic N.A.

N.A.

q 1, 2 c.

Manual Initiation N.A.

N.A.

R 1, 2 d.

High Containment Pressure Trip N.A.

N.A.

R 1, 2

((

8 n

O

~

-EMERGENCY CORE COOLING SYSTEMS I

SURVEILLANCE REQUIREMENTS _(continued)-

2.. 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:

i Valve Number

. Valve Function

a. SI-MOV-4

- LPSI Purtp Cross Over to HPSI' Pump

b. SI-MOV-22 SI Header Isolation to Cold Leg

c. SI-MOV-23 SI Header ~ Isolation to Cold Leg

- d. SI-MOV-24 SI Header Isolation to Cold Leg

e. SI-MOV-25 SI Header Isolation to Cold Leg

- l.

'3.

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-536

b. CS-MOV-537

. SI Header Isolation to Ccid Leg

' SI Header Isolation to Cold Leg i

c. CS-MOV-538 SI Header' Isolation to Cold Leg

d. CS-MOV-539 SI Header Isolation to Cold Leg i

e. MC-MOV-301 MCS Loop Isolation

f. MC-MOV-302*

MCS Loop Isolation

8. MC-MOV-309 MCS Loop Isolation

h. MC-MOV-310*

MCS. Loop Isolation

1. MC-MOV-318*

MCS Loop Isolation

j. MC-MOV-319 MCS Loop Isolation

k. MC-MOV-325 MCS Loop Isolation

1. MC-MOV-326*

MCS Loop Isolation j

1 1

i i

l i

In MODE 2, 3", 4*, 5*, power cables may be connected to the MCS loops isolation valves when required to close the valves for main coolant pump (s)-

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

i

. YANKEE-ROWE 3/4 5-5 Amendment.No. 49, 52, 69 f.

l t

w r-

-=re--.r,

--m

,,*%-=-wr*-----e e---,--4.-

e e w-rw-,------..,---r-<--

--,.--v-----...,--,.

< -. - - - -.---m

e a

~

EMERGENCY CORE COOLING SYSTEMS SL*RVEILLANCE REQUIREMENTS (cc,ntinued) 4.

Verifying that-the followin'g valves are in the.1 normally closed position with power to the valve operator re=o'.d by disconnecting the power cables as they leave the motor starter:

Valve Number Valve Function

a. CS-MOV-532 LPSI Recirculation Line

~

b. CS-MOV-534 LPSI Pump Header Isolation Valve Bypass Note: CS-MOV-532 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.

5.

Verifying that the following valves are in their normal position with power to the valve operator motors separated by dual-contactors

.l froa the motor control center:

No rmal Valve Number Valve Function Position

a. CS-MOV-533 LPSI Pump Header Isolation Open

b. CS-MOV-535 LPSI Pump Header Isolation Open

c. SI-MOV-518 LPSI Pump Header Suction Isolation Open

d. SI-MOV-48 HPSI and LPSI Minimum Recirculation Line Open

e. SI-MOV-49 HPSI and LPSI Minimus Recirculation Line Open

f. SI-MOV-515 Hot Leg Injection Isolation Closed

g. SI-MOV-514 Hot Leg Injection Isolation Closed

h. SI-MOV-516 V.C. Sump Isolation Closed

i. SI-MOV-517 V.C. Sump Isolation Closed

j. SI-MOV-46 HPSI Flow Control Open l

6.

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

7.

Verifying that each pair of ECCS recirculation subsystem redundant valves is aligned to receive elect,rical power from separate OPERABLE busses.

v 8.

Verifying that each pair of ECCS long-term hot leg injection subsystem redundant valves is aligned to receive electrical power from separate OPERABLE busses.

1 YANKEE-ROVE 3/4 5-6

' 11 Amendment No. 49, 52, 50

TABLE 3.6-1 CONTAINMENT ISOLATION VALVES T

=

li2 TESTABLE DURING VALVE NUMBER FUNCTION PLANT OPERATION ISOLATION TIME (Yes or No)

Seconds A.

AUTOMATIC ISOLATION VALVE TV-401A No. 1 SG Blowdown Yec 30 TV-401B No. 2 SG Blowdown Yes 30 TV-401C No. 3 SG Blowdown Yes 30 TV-401D No. 4 SG Blowdown Yes 30 TV-408 Containment Cooling Water Return Yes 30 TV-409 Containment Heater Condensate Return Yes 30 m

VD-S0V-301 Air Particulate Monitor-in Yes 30 VD-SOV-302 Air Particulate Monitor-out Yes 30 i

HV-S0V-1 Hydrogen Vent System Yes 30 HV-S0V-2 Hydrogen Vent System Yes 30 TV-202 Main Coolant Drain Yes 30 TV-203 Main Coolant Vent Yes 30 TV-204 Valve Stem Leakoff Yes 30 -

TV-205 Component Cooling Return No 30 TV-206 Main Coolant Sample Yes 30 TV-207 Neutren Shield Tank Samp15 Yes 30 i

TV-209 Containment Drain Yes 30 TV-211 Containment Pressure Sensing Yes 30 i

TV-212 Containment Pressure Sensing Yes 30 TV-213 LP Sample Yes 30 l

.l l

TABLE 3.6-1 (Continued) g CONTAINMENT ISOLATION VALVES e

TESTABLE DURING f3 VALVE NUMBER FUNCTION PLANT OPERATION ISOLATION TIME l

(Yes or No)

Seconds l

A.

AUTOMATIC ISOLATION VALVE (Continued)

TV-406*

Main Steam Drain to Condenser No 30 TV-411*

Atmospheric Steam Dump Ye s 30 B.

CHECK VALVES SI-V-14*

Safety Injection (IIP)

NA NA

-)

R CS-V-621*

Safety Injection (LP)

NA NA n

i Cll-V-611*

MC Feed to Loop #4 NA NA U

CC-V-667*

Component Cooling to MCP #1 NA NA

. g>

CC-V-663*

Component Cooling to MCP #2 NA NA B

CC-V-671*

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

Component Cooling to MCP #4 NA NA

~$

r*

CC-V-649*

Component Cooling to Sample Cooler Ni NA g

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 a

TABLE 3.6-1 (Cont'

'd_J, CONTAINMENT ISOLATi0h.ALVES 7.

{M TESTABLE DURING lI VALVE NilMBER FUNCTION PLANT OPERATION ISOLATION TIME p;

(Yes or No)

(Seconds)

I'I B.

CllECK VALVLS (Continued)

SW-V-820*

Service Water to Containment 7

Cooler #1 NA NA SW-V-821*

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

Service Water to Containment Cooler #3 NA NA SW-V-823*

Service Water to Containment Cooler #4 NA NA IIC-V-1199*

Steam Supply to Containment lleaters NA NA

[

C.

Manual Valves NA L

SC-MOV-551+553*

Shutdown Cooling - In No NA SC-MOV-552*S54*

Shutdown Cooling - Out No NA Cil-H0V-522*

MC Feed to Loop Fill Header NA NA CS-V-601 Shield Tank Cavity Fill NA NA CA-V-74 6 *-

Containment Air Charge NA NA IIV-V-5*

Containment 112 Vent System NA NA IIV-V-6*

Containment 112 Vent System NA NA CA-V-688 Containment 112 Vent System Air Supply NA NA CS-MOV-500 Fuel Chute Lock Valve No NA ihot subject to Type C tests

g TABLE 3.6-1 (Continued)

CONTAINMENT ISOLATION VALVES 5

TESTABLE DURING N

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 Pw p Discharge VD-V-752*

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

Neutron Shield Tank-Inner Test NA NA BF-V-4-1 Air Purge Inlet NA NA gg BF-V-4-2 Air Purge Outlet NA NA HC-V-602 Air Purge Bypass NA NA SI-MOV-516 ECCS Recirculatii.u 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

ci NRV-405B*

Main Steam Non-Return Valve No 5

NRV-405C*

Main Steam Non-Return Valve No 5

IS NRV-405D*

Main Steam Non-Return Valve No 5

Not subject to Type C tests.

q J

l l

TABLE 3.6-1 (Continued)

CONTAINMENT ISOLATION VALVES TESTABLE DURING b

PLANT OPERATION ISOLATION TIME FUNCTION b

VALVE NUMBER (Yes or No)

Seconds C.

MANUAL VALVES (Cont'd)

NA NA Main Coolant lleise Pressure Cauge PR-V-610 Purification System Containment NA PU-V-543 tR Sump Suction Purification System Containment NA PU-V-544 NA Sump Suction NA NA EBF-MOV-557*

Alternate S.G. Feed y

NA NA 3

MS-V-627***

Main Steam Bypass NA NA MS-V-628***

Main Steam Bypass NA tu 4

Main Steam Bypass NA MS-V-629***

NA y

Main Steam Bypass MS-V-630***

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

NA AS-V-720*

Steam Drain g

$a Not subject to Type C rests.

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

Not subject to type C tests.

e s

T

is PLANT SYSTEMS EMERGENCY F'

'\\TER SYSTEM.

LIMITING C0hutTION FOR OPERATION 3.7.1.2 At least two independent emergency feedwater pumps and associated flow paths shall be OPERABLE with:

-One feet ster 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 nuep 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 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

l l

l YANKEE-ROW.

3/4 7-3 Amendment No. 60, 69

e e

PLANT SYCTEMS SURVEILLANCE REQUIREMENT?,

^ 7.1.2 Each emergency feedwater pump shall be-demonstrated OPERABLE:

I a.

At least once per 31 days by:

i 1.

Starting the pump.

j s

2.

Verifying that the motot driven pump develops a discharge pressure of 2;950 psig w.

.e reci.c21ating back to.the supply tanks.

3.

Verifying that, on recirculation flow, the steam turbine driven pump develops a 12scharge pressure of 2;950 psig when the secondary steam pressure is greater than 100.psig.

4.

Verifying that the pump operates for at least 15 minutes.

5.

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

6.

Verifying that each valve in the flow path that could interrupt all emergency feedwater flow is locked open and the remaining valves are verified to be in the correct position.

b.

At least once per 10 months during shutdown by:

1.

Cycling each manual valve in the flow path that is not testable during plant operation, through at least one complete cycle of full travel.

2.

Verifying that tne steam turbine drive pump develops a discharge pressure of 2;950 psig at a flow of 2;80 gpa while feeding a steam generator and that the motor driven pumps each develop a discharge pressure of 2;950 psig with a flow of at least 80 gpa while feeding a steam generator.

3.

Cycling each main feed control valve manually through at least one complete cycle of full travel.

l c.

Prior to startup from COLD SHUTDOWN by:

1.

Verifying that each valve in the flow path from the emergency feedwater sources to the main feedwater and the steam generator blowdown lines is properly aligned to provide an uninterrupted flow path to the steam generators from the emergency feedwater system, and 2.

Performing a flow test from the emergency feedwater sources to the steam generators to verify the normal flow path.

YANKEE-ROWE 3/4 7-Sa i

Amendment No. 60, 69 l

PLANT SYSTEMS

_M_lN STEAM NON-RETURN VALVES LIMITING CONDITION FOR OPERATION 3.7.1.5 Each main steam non-return valve shall be OPERABLE.

APPLICABILITY: MODES 1, 2 and.'

l f

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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; 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 .

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 inoperpSle 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 REOUIRE'4Eh7S 4.7.1.5 Each main steam non-recurn valve that is open shall be demonstrated l

OPERABLE hy:

a.

Cycling each valve through at least 10% of full travel at least once per 92 days, and b.

Verifying full closure within 5 seconds on any closure actuation l

signal wheneve 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.

YANKEE ROWE 3/4 7-9 Amendment No. 49', 69

'l 3/4.1 REACTIVITY CONTROL SYSTEMS q

BASES t

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 subcritical from all operating conditions, 2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and 3) the reactor will le maintained sufficiently suberitical to preclude inadvertent criticality in the shutdown condition.

4 SHUTDOWN MARGIN ' requirements vary throughout core life as a function of fuel depletion, Main Coolant System boron concentration, and Main Coolant i

System T The most restrictive condition occurs at EOL, with T at noload$p$r.atingtemperature,'andisassociatedwithapostulated*sEeam line break accident and resulting uncontrolled Main Coolant System cooldown.

In the analysis of this accident, i minimum. SHUTDOWN MARGIN of 4.72% Ak/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 available at the plant. With T

< 330 F,'the reactivity transients resulting from a postulated steam 1kn! break cooldown are minimal. 5% Ak/k SHUTDOWN MARGIN (with all rods inserted) provides 3

adequate protection to preclude criticality for all postulated accidents 4

with the reactor vessel head in place.

To eliminate possible errors in the calculations of the initial reactivity of the core and the reactivity depletion rate, the predicted i

relation between fuel burnup and the boron concentration, necessary to maintain adequate control characteristics, must be adjusted (normalized) i to accurately reflect actual core conditions. Normally, when full power is reached after each refueling, and with the control rod groups in the 4

i 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 d after l

that predicted. This process of normalization should be complete 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.

1 1

1 t

YANKEE-ROWE B 3/4 1-1 Anendment No. 69 I

i

3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1.3 BORON DILUTION A minimum flow rate of at least 950 GPM provides adequate mixing, prevents stratification and ensures that reactivity changes will be gradual during boron concentration reductions in the Main Coolant Sys-tem. A flow rate of at least 950 GPM will circ'ulate an equivalent Main Coolant System volume of 2.940 cubic feet in approximately 30 minutes.

The reactivity change rate associated with boron reductions will there-fore be within the capability for operator recognition and control.

Restriction of baron dilution with Main Coolant System temperature

< 250*F prevents inadvertent criticality due to excess dilution below the temperature limit for criticality.

3/4.1.1.4 MODERATOR TEMPERATURE COEFFICIENT The limitations on moderator temperature coefficient (KTC) are provided to ensure that the assumptions used in the accident and tran-sient analyses remain valid through each fuel cycle. The surveillance requirements for measurement of the MTC during each fuel cycle are adequate to confirm the MTC value since this coefficient changes slowly due principally to the reduction in Main Coolant System boron concen-tration associated with fuel burnup. The confirmation that the measured-and appropriately compensated MTC value is within the allowable tolerance of the predicted value provides additional assurances that the coefficient will be maintained within its limits during intervals between measurement.

YANKEE-ROWE B 3/4 1-2

A 34% DNBR credit iL needed to offset the full-closure rod bow penalty in Yankee Rowe. The full-closure penalty was previously approved (D. Ross and D. Eisenhut memorandum of December 12,1976) for Yankee Rowe since a gap closure model was not available.

Generic credits (D. Edwards letter to NRC dated February 9,1977) equivalent to 13.2% DNBR margin were approved for Yankee Rowe.

The limiting transient for Yankee Rowe with respect to DNB is the 2 of 4 pump loss of flow. Based on design conditions, this event results in a minimum DNBR in excess of 2.05.

Thus, 36.6% margin to a DNBR of 1.3 exists for this limiting event, which is applied to the remaining 20.8% margin required by the rod-bow penalty.

B3/4 2-4 Amendment No. 69

'I:':*... L.. ? ~ c l --

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j 3/4.7 PLANT SYSTEMS BASES a

3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line code safety valves ensure's that the secondary system pressure will be limited to within its design pressure of 1035 psig during the.most severe anticipated system opera-

~

tional transient. The maximum relieving capacity is associated with a turbine trip from 100% RATED THERMAL POWER coincident witn an assumed loss of condenser heat sink (i.e., no steam bypass to the condenser).

The specified valve lift settings and relieving capacities are in accordance with the requirements of Section VIII of the ASP.E Boiler and Pressure Code, 1956 Edition. The total religving capacity for all valves on all of the steam lines is 3.1 x 10 1bs/hr which.is 129 percent e

6 of the total secondary steam flow of 2.4 x 10 lbs/hr at 100% RATED THERMAL POWER. A minimum of 2 OPERABLE safety valves per OPERABLE steam generator ensures that sufficient relieving capacity is available for 3'

the allowable THERMAL POWER restriction in Tables 3.7-1 and 3.7-2.

STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduction in secondary system steam flow and THERMAL POWER required by the reduced reactor trip settings of the Power Range Neutron l Flux channels. The reactor trip setpoint reductions are derived on the following bases:

For 4 loop operation SP = (X) - (Y)(V) x (108)

X For 2 loop operation SP = I*I { ( MU) x (81)

Whe e:

SP = reduced reactor trip set:cint in percent cf RATEJ THERMAL POWER J

V =.aximum number of in'operacle safety valves per stea-(

generater l ' R '.E E : 7..' E B 3/4 7 i Amendre-t';:.jyr,69

PLA}rf SYSTEMS BASES U = Maximum number of inoperable safety valves per operating steam generator (108) = Power Range and Intermediate Power Range Neutron Flux-High Trip Setpoint for 4 loop operation (81) = Maximum percent of RATED THERMAL POWER permissible for 3 loop operation X = Total relieving capacity of all safety valves per steam generator in lbs/ hour Y = Maximum relieving capacity of any one safety valve in lbs/ hour 3 /4.7.1. 2 EMERGENCY FEEDWATER SYSTEM The OPERABILITY of the emergency feedwater system ensures that the Main Coolant System can be cooled down to less than 330 F from normal operating conditions in the event of a total' loss of off-site power.

Each emergency feedwater pump is capable of delivering a total feedwater flow of 80 gpa at a pressure of 950 psig. This capacity is sufficient to ensure that adequate feedwater flow is available to remove decay heat and reduce the Main Coolant System temperature to less than 330*F when the Shutdown Cooling System may be placed into operation.

The monthly testing interval of the steam generator emergency feedwater pumps verifies their operability by recirculating water to'the supply tank. Proper functioning of the emergency feedwater pumps will be made by direct visual observation.

3/4.7.1.3 PRIMARY AND DEMINERALIZED WATER STORAGE TANK The OPERABILITY of the primary and demineralized water storage tanks with the minimum combined water volume ensures that sufficient water is available to maintain the Main Coolant System at HOT STANDBY in excess of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months with steam discharge to the atmosphere concurrent with total loss of off-site power.

t t

YANKEE-ROWE B 3/4 7-2 Amendment No. 69

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. 'I-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 a:lowance 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.

LANKEE-ROWE B 3/4 7-3 Amendment No. ST, 69 i

]

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.

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

Nominal inside diameter = 125 feet.

a.

1 b.

Minimum thickness of steel shell = 7/8 inches.

c.

Net free volume - 860,000 cubic feet.

DESIGN PRESSURE AND TEMPERATURE 5.2.2 The reactor containment is designed and shall be maintained for a 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...caining 230 or 231 fuel rods clad with Zircaloy-4.

rod shall have a nominal active fuel length of 91 inches. Each fuel assembly 4

Reload fuel 1

shall contain a maximum total weight of 234 kilograms uranium.

is similar in physical design to the Core X11 EXXON fuel and shall-have a maximum enrichment of 3.5 weight percent U-235.

I l

YANKEE-ROWE 5-1 Amendme > t No. 43, 69 1

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LOW POPULATION ZONE FIGURE 5.1-2 YANKEE-ROWE 5-3
- DESIGN FEATURES CONTROL ROD 5.3.2 The reactor core shall Jcont' in 24 control ' rods.
The control ' rods shall a
contain a nominal 90 inches of absorber material. The nominal values of'this absorber material shall be 80 percent' silver,-15. percent indium.'and 5 percent cad =fum.
The silver-indium-cadmiu= control rods shall be clad with 'Inconel.
5.4 MAIN C001 ANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The Main Coolant ' System is designed and shall be maintained:
4 a.
In accordance with the code require =ents 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 B16.5, 1957' Edition, with allowance for-3 norcal degradation pursuant to the applicable Surveillance Req ui r ece nts,
b.
For a pressure of 2500 psig, and c.
For a te=perature of 6500F, except for the pressurizer which is 6680F.
VOLUME 5.4.2 - The total water and steam voluce of the Main Coolant System.is 2940 cubic feet.
5.5 METEOROLCOICAL TORER LOCATION 5.1.1 The meteorological tower shall ye located as shown in Figure 5.1-1.
5.6 FUEL STORAGE i.
CRITICALITY i -
5.6.1 The new and spent fuel storage racks,are designed and shall be maintained -with a center-to-center distance" between fuel asse=blies placed in the storage racks to ensure'a keff equivalent to -0.95 with the new or spent fuel storage areas' flooded with unborated water. The keff of -0.95 includes.
a conse rva tive allowance of 3%
k/k for uncertainties.
YM."r:EE-ROWE 5-4 j
j Amendment No. 39, 49, 63 I
J-
ja teog o,
UNITED STATES e-Po NUCLEAR REGULATORY COMMISSION.
i g,l-
/c, E -
WASHINGTON,0. C. 20555 Sg,.....,/
l SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 69 FACILITY OPERATING LICENSE NO. OPR-3 YANKEE ATOMIC ELECTRIC COMPANY YANKEE NUCLEAR POWER STATION (YANKEE-ROWE)
DOCKET NO. 50-29 i
Date: July 22,1981
i
f-
.l
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c TABLE OF CONTENTS PAGE Introduction 1
LDiscussion
-1 Evaluation:
3 A.
Nuclear Design, Control Rod. Withdrawal Transient, Rod Ejection Transient, and Physics Related Technical Specifications 3
B.
Plant Transient and Accident Analysis 6
C.
Reactor Fuel. Design 16 D.
. Thermal Hidraulic Analysis 31 E.
Main $ team Non-Return Valve Mechanical Modifications.
42 F.
Auxiliary Feedwater System Modifications
'44 G.
Electrical and Controls Evaluation 60 f
H. -Miscellaneous Changes 67 f
Summary of Findings 67 Environmental Consideration 16 8 Conclusion
'69 i
i i
1 5
.

ML20009F329

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