Text

Amend 58 to License NPF-58.Amend Removing Functions Associated W/Msl RM from TSs 2.2,3.3.1 & 3.3.2
	ML20070M972
Person / Time
Site:	Perry
Issue date:	04/22/1994
From:	Hopkins J; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20070M975	List: ML20070M972; ML20070M978;
References
	NUDOCS 9405040265
	Download: ML20070M972 (14)
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S UNITED STATES i

NUCLEAR REGULATORY COMMISSION

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WASHINGTON, D.C. 2055A001 THE CLEVELAND ELECTRIC ILLUMINATING COMPANY. ET AL.

DOCKET NO. 50-440 PERRY NUCLEAR POWER PLANT. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.58 License No. NPF-58 1.

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

A.

The application for amendment by The Cleveland Electric Illuminating Company, Centerior Service company, Duquesne Light Company, Ohio Edison Company, Pennsylvania Power Company, and Toledo Edison Company (the licensees) dated September 28, 1992, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

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

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

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.

2.

Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-58 is hereby amended to read as follows:

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4 9405040265 940422 PDR ADOCK 05000440 p

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(2)

Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 58 are hereby incorporated into this license. The Cleveland Electric Illuminating Company shall operate the facility in accordance with the Technical Specifi-cations and the Environmental Protection Plan.

3.

This license amendment is effective as of its date of issuance and shall be implemented not later than 90 days after issuance..

FOR THE NUCLEAR REGULATORY COMMISSION

/ gx

/.

Jon B. Hopkins, Sen'ior Project Manager

/,

Project Directorate III-3

' ~

Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of issuance: April 22,1994

ATTACHMENT TO LICENSE AME.NDMENT NO. 58 FACILITY OPERATING LICENSE NO. NPF-58 DOCKET NO. 50-440 Replace the following pages of the Appendix "A" Technical Specifications with the attached pages.

The revised pages are identified by Amendment number and contain vertical lines indicating the area of change. Overleaf pages are provided to maintain document completeness.

Remove Insert 2-4 2-4 B 2-8 B 2-8 3/4 3-2 3/4 3-2 3/4 3-4 3/4 3-4 3/4 3-6 3/4 3-6 3/4 3-7 3/4 3-7 3/4 3-11 3/4 3-11 3/4 3-15 3/4 3-15 3/4 3-16 3/4 3-16 3/4 3-21 3/4 3-21 3/4 3-23 3/4 3-23 3/4 3-26 3/4 3-26 l

l i

l

TABLE 2.2.1-1 Y

_ REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS m

y FUNCTIONAL UNIT ALLOWABLE TRIP SETPOINT VALUES 5

1.

Intermediate Range Monitor

-4 a.

Neutron Flux-High

< 120/125 divisions

< 122/125 divisions b.

Inoperative 5f full scale if full scale NA NA 2.

Average Power Range Monitor:

a.

Neutron Flux-High Setdown

< 15% of RATED

< 20% of RATED THERMAL POWER THERMAL POWER b.

Flow Biased Simulated Thermal Power-High

1) Flow Biased

< 0.66 W+64%, with 1 0.66 W+67%, with

2) High Flow Clamped a maximum of a maximum of

< 111.0% of RATED

< 113.0% of RATED c.

Neutron Flux-High THERMAL POWER THERMAL POWER

< 118.0% of RATED

< 120.0% of RATED d.

Inoperative THERMAL POWER THERMAL POWER WA

.NA 3.

Reactor Vessel Steam Dome Pressure - High i 1064.7 psig i 1079.7 psig 4.

Reactor Vessel Water Level - Low, level 3 2

> 177.7 inches above

> 177.1 inches above E

Top of active fuel

Top of active fuel

2

5. ' Reactor Vessel Water Level-High, level 8 P.

< 219.5 inches above

< 220.1 inches above Top of active fuel

Top of active fuel

s 6.

Main Steam Line Isolation Valve - Closure 1 8% closed i 12% closed T

7.

Deleted 8.

Drywell Pressure - High i 1.68 psig i 1.88 psig "See Bases Figure B 3/4 3-1.

w

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS l

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

APPLICABILITY: As shown in Table 3.3.1-1.

ACTION:

With a reactor protection system instrumentation setpoint less conservative than the value shown in the Allowable Values column of Table 2.2.1-1, declare l

the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1 until the channel is restored to OPERABLE status with its setpoint adjusted consistent with the Trip Setpoint value.

9 PERRY - UNIT 1 2-3

i

{,.

LIMITING SAFETY SYSTEM $ETTINGS J

j BASES J

j REACTOR PROTECTION SYSTEM INSTRtMENTATION SETPOINTS (Continued) i j

Aversoe Power Rance Monitor (Continued) 5% of RATED THERMAL POWER per minuta and the APRM system would be more than 1

adequate to assure shutdown before the power could exceed the Safety Limit.

The 15% neutron flux trip remains active until the mode switch is placed in 2

i j

the Run position.

The APRM trip system is calibrated using heat balance data taken during steady state conditions.

Fission chambers provide the basic input to the sys-tem and therefore the monitors respond directly and quickly to changes due to i

transient operation for the case of the Neutron Flux-High setpoint; i.e, for a power increase, the THERMAL POWER of the fuel will be less than I

that indicated by the neutron flux due to the time constants of the heat trans-i for associated with the fuel.

For the Flow Biased Simulated Thermal Power-High setpoint, a time constant specified in the COLR is introduced into the flow a

g i

biased APRM in order to simulate the fuel thermal transient characteristics.

A more conservative maximum value is used for the flow biased setpoint as shown j

in Table 2.2.1-1.

i The APRM setpoints were selected to provide adequate margin for the Safet 4

Limits and yet allow operating margin that reduces the possibility of unneces y l

sary shutdown.

j 3.

Reactor Vessel Steam Dome Pressure-High i

High pressure in the nuclear system could cause a rupture to the nuclear system process barrier resulting in the release of fission products. A pres-sure increase while operating will also tend to increase the power of the reactor by compressing voids thus adding reactivity.

The tri reduce the neutron flux, counteracting the pressure increase.p will quickly The trip set-j ting is slightly higher than the operating pressure to permit normal operation without spurious trips.

The setting provides for a wide margin to the maximum j

allowable design pressure and takes into account the location of the pressure i

measurement compared to the highest pressure that occurs in the system during i

a transient.

This trip setpoint is effective at low power / flow conditions when i

the turbine control valve fast closure and turbine stop valve closure trips are i

bypassed.

For a load rejection or turbine trip under these conditions, the.

j transient analysis indicated an adequate margin to the thermal Aydraulic limit.

l i

i I

PERRY - UNIT 1 8 2-7 Amendment No. 48 Ult / 2 8 E:t 1

LIMITING SAFETY SYSTEM SETTINGS BASfS

^

9 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued) 4.

Reactor Vessel Water Level-Low The reactor vessel water level trip setpoint has been used in transient analyses dealing with coolant inventory decrease. The scram setting was chosen far enough below the normal operating level to avoid spurious trips but high enough above the fuel to assure that there is adequate protection for the fuel and pressure limits.

5.

Reactor Vessel Water Level-Hiah A reactor scram from high reactor water level, approximately two feet above normal operating level is intended to offset.the addition of rsactivity effect associated with the introduction of.a significant amount of relatively cold feedwater. An excess of feedwater entering the vessel would be detected by the level increase in. a timely manner.

This scram feature is only effective when the reactor mode switch is in the Run positionbecause at THERMAL POWER levels below 10% to 15% of RATED THERMAL POWER, the approximate range of' power level for changing to the Run position, the safety margins are more than adequate without a reactor scram.

6.

Main Steam line Isolation Valve-Closure The main steam line isolation valve closure. trip was provided to limit 1

the amount of fission product release for certain postulated events. The MSIV's are closed automatically from measured parameters such as high steam flow, low reactor water level, high steam tunnel temperature and low steam line pressure.

The MSIV's closure scram anticipates the pressure and flux transients which could follow MSIV closure and thereby protects reactor vessel pressure and fuel thermal / hydraulic Safety Limits.

7.

Deleted

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PERRY - UNIT 1 B 2-8 Amendment No.58

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f 3/4.3 INSTRUMENTATION 1

i 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION j

LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the reactor protection system instrumentation channels shown in Table 3.3.1-1 shall be OPERABLE with the REACTOR PROTECTION SYSTEM j

RESPONSE TIME as shown in Table 3.3.1-2.

l APP'.ICABILITY: As shown in Table 3.3.1-1.

ACTION:

With the number of OPERABLE channels less than required by the Minimum a.

OPERABLE Channels per Trip System requirement for one trip system, place the inoperable' channel (s) and/or that trip system in the tripped condition

within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . The provisions of Specification 3.0.4 are not applicable.

b.

With the number of OPERABLE channels less'than required by the Minimum OPERABLE Channels per Trip System requirement for both trip systems, place at least one trip system ** in the tripped condition within ont hour and j

take the ACTION required by Table 3.3.1-1.

SURVEILLANCE REQUIREMENTS I

4.3.1.1 Each reactor protection system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL' CHECK, CHANNEL FUNCTIONAL i

TEST and CHANNEL CALIBRATIO'1 operations for the OPERATIONAL CONDITIONS and at j

the frequencies shown in Teole 4.3.1.1-1.

4.3.1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of

{

all channels shall be performed at least once per 18 months.

4.3.1.3 The REACTOR PROTECTION SYSTEM RESPONSE TIME of each reactor trip 2

functional unit shown in Table 3.3.1-2 shall be demonstrated -to be within its limit at least once per 18 months.

Each test shall include at least one i

j channel per trip system such that all channels are tested at least once every l

i N times 18 months where N is the total number of redundant channels in a specific reactor trip system.

4.3.1.4 The provisions of Specification 4.0.4 are not applicable to the CHANNEL j

FUNCTIONAL TEST and CHANNEL CALIBRATION surveillances for the Intemediate Range Monitors for entry into their applicable OPERATIONAL CONDITIONS (as shown in Table 4.3.1.1-1) from OPERATIONAL CONDITION 1, provided the surveillances are i

performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after such entry.

I

An inoperable channel need not bEplaced in the tripoed condition where this would cause the Trip Function to occur.

In these cases, the inoperable channel 1

shall be restored to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or the ACTION required by Table 3.3.1-1 for that Trip Function shall be taken.

- The trip system need not be placed in the tripped ~ondition if this would causi '.

c the Trip Function to occur. When a trip system can be placed in the tripped i

condition without c'ausing the Trip Function to occur, place the trip system with the most inoperable channels in the tripped tondition; if both systems

~

have the same number of inoperable channels, place either trip system in the tripped condition.

PERRY - UNIT 1 3/4 3-1 Amendment No. 41 3

MAR 2 01392 c

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.--..---.. i*.,.

m m

IABLE 3.3.J-1 REACTOR PROTECTION SYS1EM INSTRUMENTATION-t E

APPLICABLE MINIMUM OPERATIONAL OPERABLE CHANNELS M

TUNCTIONAL UNIT CONDITIONS PER TRIP SYSTEN (31 ACTION t

a

.l.

Intermediate Range Monitors:

a.

Neutron Flux - High 2

3 1

~

3(b)4 3

2 5

3 3

1 b.

Inoperative 2

3 1

1 3, 4 3

2 5

3 3

2.

Average Pc wr Range Monitor '*):

a.

Neuton Clux - High, Setdown 2

3 1

3

-3 R

5(b) 2 3

3 b.

Flow Blased Simulated Thermal 4

Power - High 1

3 4

.c.

Neutron Flux - High 1

3 4

d.

Inoperative 1, 2 3

1

[

3 2

3 5-o 3

3 E

3.

Reactor Vessel Steam Dome

'g*

Pressure - High I, 2'*

2 1

Ei 4.

Reactor Vessel Water Level - Low, g

level 3 1, 2 2

1 5.

Reactor. Vessel Water Level - High, Level 8 l

2 4

6.

Main Steam Line Isolation Valve -

Closure 1(*)

4 4

7.

Deleted y

8.

Drywell Pressure - High

1. 2")

i

?

. t

TABLE 3.3.1-1 (Continued) h REACTOR PROTECTION SYSTEM INSTRUMENTATION N

e APPLICABLE MINIMUM OPERATIONAL OPERABLE CHANNELS

[

FUNCTIONAL UNIT CONDITIONS PER TRIP SYSTEM (a)

ACTION 9.

Scram Discharge Volume Water Level - High a.

Level Transmitter 1, 2 2

1 I9) i S

2 3

l b.

Float Switches 1, 2 2

1 I9)

S 2

3 Ih) 10.

Turbine Stop Valve - Closure I

4 6

11.

Turbine Control Valve Fast Closure, Valve Trip System 011 Pressure - Low 1(h) w 2

6 12.

Reactor Mode Switch Shutdown Position 1, 2 2

1-3, 4 2

7

)

5 2

3 13t Manual Scram 1, 2

'2 1

3, 4 2

8 5

2 9

i

\\

t

TABLE 3.3.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION ACTION ACTION 1 -

Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 2 -

Verify all insertable control. rods to be inserted in the core and lock the reactor mode swithch in the Shutdown position within one hour.

ACTION 3 -

Suspend all operations involving CORE ALTERATIONS

and insert all insertable control rods within one hour.

ACTION 4 -

Be in at least STARTUP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 5 -

Deleted.

ACTION 6 -

Initiate a reduction in THERMAL POWER within 15 minutes and reduce turbire first stage pressure to less than the automatic bypass setpoint yithin 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

ACTION 7 -

Verify all insertable control rods to be inserted within one hour.

ACTION 8 -

Lock the reactor mode switch in the Shutdown position within one hour.

ACTION 9 -

Suspend all operations involving CORE ALTERATIONS

and insert all insertable control rods and lock the reactor mode switch in the Shutdown position within one hour.

'Except replacement of LPRM strings provided SRM instrumentation is OPERABLE per Specification 3.9.2.

PERRY - UNIT 1 3/4 3-4 Amendment No38

TABLE 3.3.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRLMENTATION TABLE NOTATIONS (a) A channel say be placed in an inoperable status for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter.;

(b) Unless adequate shutdown margin has been demonstrated per Specifica-tion 3.1.1 and the "one-rod-out" Refuel position interlock has been demonstrated CPERABLE per specification 3.9.1, the shorting links shall be removed from the RPS circuitry prior to and during the time any control rod is withdrawn."

(c) An APRM channel is inoperable if there are less than 2 LPRM inputs per level or less than 14 LPRM inputs to an APRM channel.

(d) This function is not required to be OPERABLE when the reactor pressure vessel head is removed per Specification 3.10.1.

(e) This function shall be automatically bypassed when the reactor mode switch is not in the Run position.

(f) This function is not required to be OPERABLE when DRYWELL INTEGRITY is not required.

(g) With any control rod withdrawn.

Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

(h) This function is automatically bypassed when turbine first stage pressure is less than the value of turbine first stage pressure corresponding to 40%** of RATED THERMAL POWER.

Not required for control rods removed per Specification 3.9.10.1 or 3,9.10.2.

- The Turbine First Stage Pressure Bypass Setpointb and corresponding AllowableValuesareadjustedbasedonFeedwatertemperatures(see3/4.2.2 for definition of AT). The Setpoints and Allowable Values for vatrious ATs are as follows:

T('F)

Setpoint (osic)

Allowable Value (osic) 0=T 1 212 1 218 1 190 1

1 96

) < AT < 50 50 < AT < 100 1 168 1 174 100 < AT 1 170 1 146 1 152

i pIRRY - UNIT 1 3/4 3-5 Amendment No. 29

,3 E

TABLE 3.3.1-2 a_,s REACTOR PROTECTION SYSTEM RESPONSE TIMES t

E

~~

-4 FUNCTIONAL UNIT RESPONSE TIME (Secondsl l.

Intermediate Range Monitors:

Neutron Flux - High a.

b.

Inoperative NA NA 2.

~ Average Power Range Monitor *:

a.

Neutron Flux - High, Setdown NA b.

Flow Biased Simulated Thermal Power - High 5 0.09'*

Neutron Flux - High c.

d.

Inoperative 5 0.09 NA 3.

Reactor Vessel Steam Dome Pressure - High 4.

Reactor Vessel Water Level - Low, level 3 s 0.35 u,

];

5.

Reactor Vessel Water Level - High, level 8 5 1.05 6.

Main Steam Line Isolation Valve - Closure s 1.05 u,

f, 7.

Deleted s 0.06 8.

Drywell Pressure - High l

9.

' Scram Discharge Volume Water Level - High NA 10.

Turbine Stop Valve - Closure NA 11.

Turbine Control Valve Fast Closure, Valve Trip System 5 0.06 011 Pressure - Low

12. Reactor Mode Switch Shutdown Position s 0.07#

!i

13. Manual Scram NA h

NA

' Neutron detectors are exempt from response time testing.

2e the detector output or from the input of the first electronic component in the channel. Response time s P

im

..Not including the simulated thermal power ti Fa me constant specified in the COLR.

E

Measured from start of turbine control valve fast closure.

+

e

y TABLE 4.3.1.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS' I

c5 CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH

[

FUNCTIONAL UNIT CHECK TEST' CALIBRATION (*)

SURVEILLANCE REQUIRED 1.

Intermediate Range Monitors:

a.

Neutron Flux - High

-5/U,5,(b)

W R

2 S

W.

R 3, 4 b.

Inoperative NA W

NA 2, 3, 5

2.

Average Power Range Monitor:(f) a.

Neutron Flux - High, S/U,5,(b)

W SA 2

Setdown S

W SA 3, 5 b.

Flow Biased Simulated Thermal Power - Hig".

S,0(h) y g(d)(e)

SA(*), R(')

1 y

c.

Neutron Flux - High S

W W(d', SA 1

w d.

Inoperative

.NA W

NA 1,.2, 3, 5 3.

Reactor Vessel-Steam Dome Pressure - High S'

M R(8) 1, 2 '

0 4.

Reactor Vessel Water Level -

Low, level 3

.S M

R(8)

~ 1, 2

,a E

5.

Reactor Vessel Water Level -

y High, level 8 S

M-R(

I

=>

6.

Main Steam Line Isolation f

Valve - Closure NA M~

R I

7.

Deleted I

8.

Drywell Pressure - High S

.M R(')

' 1, 2")

9.

Scram Discharge Volume Water

. Level - High i

a.-

Level' Transmitter S

M R(')

'l.- 2, 5(*)

b.

Float Switches NA M'

R.

1. 2, 5(*)

l

\\

j TABLE 4.3.1.1-1 (Continued)

R E

REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL i

CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WICH i

E FUNCTIONAL UNIT CHECK TEST Call 8 RATION SURVEILLANCE REQUIRED U

10. Turbine Stop Valve - Closure NA M

R 1

wa.

11. Turbine Control Valve fast closure Valve Trip System 011 Pressure - Low NA M

R 1

12.

Reactor Mode Switch a

Shutdown Position MA R

NA 1,2,3,4,5

13. Manual Scram NA M

MA 1,2,3,4,5 M neutron detectors any be excluded from CHANNEL CALIBRATION.

(b)

The Im and Sm channels shall be determined to overlap for at least 1/2 decades during each startup i

after entering OPERATIONAL CONDITION 2 and the IRM and APRM channels shall be determined to overlap for i

t' at least 1/2 decades during each controlled shutdown, if not performed within the previous 7 days.

i (c)

Deleted i

T (d) This calibration'shall consist of the adjustment of the APRM c'annel to confers to the power values h

calculated by a heat balance during OPERATIONAL CONDITION 1 wnen THERMAL POWER > 255 of RATED THERMAL POWER.

i Adjust the AP M channel if the absolute difference is greater than 25 of RATED THERMAL POWER.

The provisions of Specification 4.0.4 are not applicable provided the surveillance is. performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reaching 25% of RATED THERMAL POWER.

(e)

This calibration shall consist of the adjustment of the APM flow biased channel to conform to a calibrated flow signal.

l (f)

The LPRMs shall be calibrated at least once per 1000 W D/T using the TIP system.

(g)

Calibrate trip unit setpoint at least once per 31 days.

i l

y (h) existing loop flow (APRM X flow). Verify measured core flow (total core flow) to be greater than or a

y (1)

This calibration shall consist of verifying that the simulated thermal power time constant is within r;.

the limits specified in the COLR.

ir3

'(j) This function is not required to be OPERA 8LE when the reactor pressure vessel. head is removed l om per Specification 3.10.1.

*?

(k) With any control rod withdrawn. Not applicable to control rods removed per

. rf 3 Specification 3.9.10.1 or 3.9.10.2.

ib-(1) This function is not required to be OPERA 8LE when Drywell Integrity is not required.

.(a) The CHANNEL CALIBRATION shall exclude the flow reference transmitters, these transmitters shall be l

calibrated at least once per 18 months.

l i

l 1

I

a TABLE 3.3.2-1 ISOLATION ACTUATION' INSTRUMENTATION g

MINIMUM APPLICABLE

=

OPERABLE CHANNELS OPERATIONAL-

'7 JRIP FUNCTION PER TRIP SYSTEM (al

' CONDITION ACTION g

1.

PRIMARY CONTAINMENT ISOLATION Z

a.

Reactor Vessel Water Level -

Low, level 2 (Division I & 2) 2 1, 2, 3 and f 20

~

b.

Reactor Vessel Water Level -

Low, level 2 (Division 3) 4"'

1, 2, 3 and #

28 c.

Drywell Pressure - High (Division l'& 2) 2 1, 2, 3.

20 d.

Drywell' Pressure - High (Division 3) 4 "'

1, 2, 3 28 e.

Containment and Drywell Purge Exhaust Plenum Radiation - High 2(b) 1, 2, 3 and

21 f.

Reactor = Vessel Water Level -

R Low, level 1 2

1, 2, 3 and #

20 g.

Manual Initiation y

(Division I & 2) 2"'

1, 2, 3 and *L 22 h.

Manual Initiation (Division 3) 1")

1, 2, 3 and

28

~

2.

MAIN STEAM ISOLATION a.

Reactor. Vessel. Water Level -

Low, Level 1 2

1, 2, 3 20 g

b.

Main-Steam Line a

Radiation - High 2")

29 g

c.

Main Steam Line g

Pressure-- Low 2

1-24 d.-

Main Steam Line 2

P Flow - High 2/1ine

.1, - Z 3 23 u

w e.

Condenser Vacuum - Low.

2 1,2

,3,,

23

?

f.

Main Steam Line Tunnel w

Temperature - High 2-1, 2,'3

23

?

g.

Main Steam Line Tunnel 8;

A Temperature - High 2

1, 2, 3 23 h.

. Turbine Building Main Steam Line Temperature - High 2

1, 2, 3 23

- 1.

Manual Initiation 2

-1, 2, 3 22

l e

l E'

4 2 C 22 22 2

2 2

2 222 i

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1 5

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=

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R t**

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Ag 5

..I*A I. t= g E

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E stj 5

' tu s s5 -3 E naje a g e iil1j;g)jles 3e i

a

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=-

a eel C C '5 '

3 s8i e 4 bE b

$' E$ E Md 44 5444 W

4 u

4 4.:

k ci 4

PERRY - LNIT 1 3/4 3 12 Amendment No. 44 t

DCT2 81?]2

j TABLE 3.3.2-1 (Continued) l ISOLATION ACTUATION INSTRUMENTATION j

~

ACTION ACTION 20 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN w In OPERATIONAL CONDITION f, suspend CORE ALTERATIONS and i

operations with a potential for draining the reactor vessel.

i ACTION 21 Close the affected system isolation valve (s) within one hour or:

In OPERATIONAL CONDITION 1, 2 or 3 s.

SHUTDOWN within the next 12. hours a be in at least HO nd in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

1 b.

In Operational Condition *, suspend CORE ALTERATIONS handling of irradiated fuel in the primary containmen,t and l

operations with a potential for. draining the reactor vessel.

ACTION 22 Restore the manual initiation function to CPERABLE statu within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or:

t

-In OPERATIONAL CONDITION 1, 2, or 3, be in at least HOT a.

SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHU within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

i b.

In OPERATIONAL CONDITION *, suspend CORE ALTERATION operations with a potential for draining the reactor i

vessel, and handling of irradiated fuel in the primary containment.

i ACTION 23 Se in at least STARTUP with the associated isolation' va closed within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in at least HOT SHUTDOWN 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 ho I

ACTION 24 Se in at least STARTUP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

)

i ACTION 25 Verify SECONDARY CONTAINMENT INTEGRITY with the annulu gas treatment system operating within one hour.

ACTION 26 Restore the manual initiation function to OPERABLE sta within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or close the affected system' isolation valves within I hour and declare the affected system inoperable.

I ACTION 27 Close the affected system isolation valves within' one hour and declare the affected system inoperable.

ACTION 28 Within one hour lock the affected system isolation valves close or verify, by remote indication, that the valve (s) is closed and electrically disarmed, or isolate the penetration (s) and declare the affected system inoperable.

ACTION 29 Close the associated isolation valves within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

MIES When handling irradiated fuel in the primary contairvaent ALTERATIO C and operations with a potential for draining th and during CORE Condenser Low Vacuum Bypass Switch is in reactor vessel.

r the key locked f

During CORE ALTERATIONS and operations with a potential on.

reactor vessel.

raining the OPERATIONAL CONDITION I or 2 when the mechanical vacuum pump 7

are not' isolated.

PERRY - UNIT 1 3/4 3-15 Amendment-No.38,42,56,58-

,,9 gn,,,wm._,p,,.,_m-,,_,,,,.%pppey,q.w.,.

~. _..

m

,y.y, y,,,v y,

I

~

. TABLE 3.3.2-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION ACTION NOTES (Continued)

(a)

A channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for condition provided at least one other OPERABLE ch trip system is monitoring that parameter.

(b)

Containment and Drywell Purge System inboard and outboard isolation valves each use a separate two out of two isolation logic.

(c)

There is only one (1) RCIC manual initiation channel for RCIC system containment isolation valves.

(d)

Division 3 has onl combined in a one y one trip system consisting of four channels logically out-of-two-twice configuration which only closes the HPCS Suppression Pool Test Return Valve (1E22-F023).

(e)

Division 3 Manual Initiation consists of a single channel in a single trip system.

(f)

This Trip Function no longer isolates the Main Steam Lines isolation is of the mechanical pump lines The only AandB),

one-out-of-two logic.

na PERRY - UNIT 1 3/4 3-16 Amendment No. #f,70,58

)'

TABLE 3.3.2-3 ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME TRIP FUNCTION RESPONSE TIME (Seconds)#

~

l.

PRIMARY CONTAINMENT ISOLATION I

Reactor Vessel Water Level - Low, Level 2 NA-a.

b.

Drywell Pressure - High NA Containmentandgr c.

Radiation - High g ell Purge Exhaust Plenum s 10(*)

]

d.

Reactor Vessel Water Level - Low, Level I NA e.

Manual Initiation NA 2.

MAIN STEAM LINE ISOLATION Reactor Vessel Water Level - Low, Level I s 1.0*/s 10(*)**

a.

b.

Main Steam Line Radiation - High NA 1.0*/s 10(*)**

l c.

Main Steam Line Pressure - Low i

d.

Main Steam Line Flow - High i 0.5 /s 10(*)**

e.

Condenser Vacuum - Low NA f.

Main Steam Line Tunnel Temperature - High NA g.

Main Steam Line Tunnel A Temperature - High NA h.

Turbine Building Main Steam Line Temperature - High NA 1.

Manual Initiation NA t

3.

SECONDARY CONTAINMENT ISOLATION 1

Reactor Vessel Water Level - Low, Level 2 NA a.

b.

Drywell Pressure - High NA c.

Manual Initiation NA l

4

4. REACTOR WATER CLEANUP SYSTEM ISOLATION a.

A Flow - High b.

A Flow Timer NA NA Equipment Area Temperature - High NA i

c.

d.

Equipment Area A Temperature - H1 h NA Reactor Vessel Water Level - Low,3 Level 2 NA e.

f.

Main Steam Line Tunnel Ambient

)

Temperature - High NA g.

Main Steam Line Tunnel A Temperature - High NA SLCS Initiation

{

h.

NA 1.

Manual Initiation NA PERRY - UNIT 1 3/4 3-21 Amendment No.58

TABLE 3.3.2-3 (Continued)

ISOLATION SYSTEM INSTRUMENTATION RESPO TRIP FUNCTION 5.

RESPONSETIME(Seconds]#

REACTOR CORE ISOLATION COOLING SYSTEM ISO 4

RCIC Steam Line Flow - High a.

b.

RCIC Steam Supply Pressure - Low NA RCIC Equipment Room' Ambient "esperatur c.

NA d.

i NA RCIC Equipment Room A Temperature - High e.

NA f.

Main Steam Line Tunnel Ambient NA Temperature - High g.

Main Steam Line Tunnel Temperature TimerMain NA h.

NA 1.

RHR Equipment Room A Temperature - HighRH NA 3

.j.

MA k.

RCIC Steam Line F NA Drywell Pressure low High Timer 1.

j_

High NA m.

Manual Initiation NA 1

~

MA 6.

RHR SYSTEM ISOLATION RHR Equipment Area A Temperature - HighRHR a.

b.

NA RHR/RCIC Steam Line Flow - High c.

1 NA d.

Reactor Vessel Water Level - Low, Level 3 NA Reactor Vessel (RHR Cut-in Permissive) e.

NA Pressure - High f.

Drywell Pressure - High NA Manual Initiation NA g.

NA (a) Isolation system instrumentation response time specifi d i generator starting and sequence loading delays.

e ncludes the diesel (b) Radiation detectors are exempt from response time testi shall be measured from detector output or the input of the fir ng.

Response time component in the channel.

onic

' Isolation system instrumentation response time for NSIVs on generator delays assumed.

No diesel

- Isolati rn system instrumentation response time for associated valves exceot Mivs.

91 solation system instrumentation response time specifi d f Function actuating each containment isolation valve shall be e

or the Trip the isolation time for each valve to obtain ISOLATIO added to TIME for each valve.

ESPONSE PIRRY - UNIT 1 3/4 3*22 Amendment No. 44 00T381M

L TABLE 4.3.2.1-1 j

ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS A

CHANNEL OPERATIONAL y

CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED l.

PRIMARY CONTAINMENT ISOLATION E

a.

Reactor Vessel Water Level -

Z Low, level 2 S

M R*)

1, 2, 3 and #

b.

Drywell Pressure - High S

M R*)

1, 2, 3 c.

Containment and Drywell Purge Exhaust Plenum Radiation -

High S

M R

d.

Reactor Vessel Water level -

1,-2, 3 and

Low, Level 1 S

M R

1, L 3 ud #

ie.

Manual Initiation NA R

NA 1, 2, 3 and

2.
MAIN STEAM LINE ISOLATION a.
Reactor Vessel Water Level -
Low, Level 1 S
M R(b) 1, 2, 3 b.
Main Steam Line Radiation -
w2 High S
M R
w c.
Main Steam Line Pressure -
l A,
Low S
M R*)
I w
d.
Main Steam Line Flow - High S
M R*)
1, 2 3 e.
Condenser Vacuum - Low S
M Rcb>
1, 2, 3**
3 f.
Main Steam Line Tunnel Temperature - High S
M R
1, 2, 3 g.
Main Steam line Tunnel A Temperature - High S
M R
1, 2, 3 y
h.
Turbine Building Main Steam g
Line Temperature - High S
M R
1, 2, 3 l.
Manual Initiation NA R
NA 1, 2, 3 k
EF i
8 i
i
TABLE 4.3.2.1-1 (Continued) h ISOLATION ACTUATION-INSTRUMENTATION SURVEILLANCE REQUIREMENTS 4
CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH
^
g TRIP FUNCTION
. CHECK TEST CALIBRATION SURVEILLANCE REQUIRED U
3.
SECONDARY CONTAINMENT ISOLATION w
a.
Reactor Vessel Water Level - Low, Level 2 S
M R(b)
' b.
Drywell Pressure - High S
M R(b) 1, 2, 3 and #
1, 2, 3 c.
Manual Initiation NA R
NA 1, 2, 3 and
4.
REACTOR WATER CLEANUP SYSTEM ISOLATION a.
A Flow - High S
M R
1,2,3 b.
A Flow Timer NA M
R c.
Equipment Area Temperature -
1,2,3 High 5-M R
d.
Equipment Area Ventilation 1, 2, 3 w}
A Temperature - High S
M R
'1, 2, 3 w
e.
~ Reactor Vessel Water 4
Level - Low,' Level 2 S
M Rg) f.
Main Steam Line Tunnel Ambient 1, 2,~ 3 Temperature - High S
M R
1, 2, 3 g.
Main Steam Line Tunnel A Temperature - High 5
M R
h.
SLCS Initiation NA M(,)
1,2,3 NA 1,2,3 i.
Manual Initiation NA R
NA 1,'2, 3 1
I i
\\
t
I S
TABLE 4.3.2.1-1 (Continued) h ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS E
CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH E
TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED U
5.
REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION a.
RCIC Steam Line Flow - High S
M
-R(b) 1, 2, 3 b.
RCIC Steam Supply Pressure -
Low S
M R(b) 1, 2, 3 c.
RCIC Turbine Exhaust Diaphragm Pressure - High S'
M R(b) 1,2,3 d.
RCIC Equipment Room Ambient Temperature - High S
M R
1, 2, 3 e.
RCIC Equipment Room a Temperature - High S
M R
1,2,3 f.
Main Steam Line Tunnel Ambient Temperature - High S
M R
1,2,3 w1 g.
Main Steam Line Tunnel A Temperature - High S
M R
1,2,3 w
J, h.
Main Steam Line Tunnel Temperature Timer NA M
R 1,2,3
~
i.
RHR Equionent Room Ambient Temperature - High S
M R
1,2,3 J.
RHR Equipment Room A Temperature - High S
M R
1,2,3 k.
RCIC Steam Line Flow NA M
R 1,2,3 4
~
High Timer 1.
Drywell Pressure - High S
M R(b)-
1, 2, 3 m.
Manual Initiation
.NA R
NA 1,2,3 1
+
r i
n I
,s 93 TABLE 4.3.2.1-1 (Continued) 35 JSOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS e
CHANNEL OPERATIONAL
_ TRIP FUNCTION CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH c-
_ CHECK TEST CALIBRATION SURVEILLANCE REQUIRED 25 6.
RHR SYSTEM ISOLATION
~4 a.
RHR Equipment Area Ambient Temperature - High S
M R
1, 2, 3 b.
RHR Equipment Area A Temperature - High S
M R
1, 2, 3 c.
RHR/RCIC Steam Line Flow - High S
M R'b' 1, 2, 3 d.
Reactor Vessel Water Level -
Low, level 3 S
M R"*
1, 2, 3 e.
Reactor Vessel (RHR Cut-in Permissive) Pressure - High S
M R "
s 1, 2, 3 f.
Drywell Pressure - High 5
M R""
w d>
1, 2, 3 g.
Manual Initiation NA R
NA 1, 2, 3 l
tr j
When handling irradiated fuel in the primary containment and during CORE ALTERATIONS Et with a potential for draining the reactor vessel.
operations
When any turbine.stop valve is greater than 90% o~ en and/or the key locked bypass s it normal-position.
p w c s in the-i EF
.# During CORE: ALTERATION (a) Each train or logic channel shall be tested at least every other 31 days.and operati i
in l
{,b)Calibratetripunitsetpointatleastonceper31 days.
OPERATIONAL CONDITION 1 or 2'when the mechanical vacuum pump lines are not-isolated 3
r i
t i
.

ML20070M972

Text

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