Text

Amend 75 to License NPF-43,replacing Listed Pages of App a Tech Specs W/Attached Pages
	ML20082U403
Person / Time
Site:	Fermi ; (NPF-43-A-075, NPF-43-A-75)
Issue date:	09/06/1991
From:	Marsh L; Office of Nuclear Reactor Regulation
To:	;
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ML20082U406	List: ML20082U403; ML20082U427;
References
	NUDOCS 9109200152
	Download: ML20082U403 (28)
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DETR0li EDISON COMPANY DOCKET NO. 50-341 FERMI-2 AMENDMENT 10 FACILITY OPERATING LICENSE Amendment No. 75 License No. NPF-43 1.

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

A.

The application for amendment by the Detroit Edison Company (the licensee) dated August 20, 1990, 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 (1) that the activities authorized by this anendment 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; D.

The issuance t.f this amendment will not be inimical to the common defens,e 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 Specifica-tions as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No NPF-43 is hereby amended to read as follows:

3chnical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A as revised through Amendment No. 75, and the Environmental Protection Plan contained in l

l A3pendix B, are hereby incorporated in the license. Deco shall operate tie facility in accordance with the Technical Specifications and the Environmental Protection Plan.

910920015J 910906 PDR ADOCK 05000341 P

PDR l

2 3.

This license amendnent is effective as of its date of issuance.

r FOR THE NUCLEAR REGULATORY COMMISSION w

<\\ L. B. Marsh, Director Project Directorate 111-1 Division of Reactor Projects 111/IV/V Office of Nuclear Reactor Regulation

Attachment:

' Charges to the Technical Specifications Date of Issuance:

September 6,1991 i

4 l

,,.,a.+

m,,.-..

ATTACHMENT TO L1 CENSE AMENDMENT NO. 75 FACILITY OPERATING LICENSE N0. NPF-43 DOCKET NO. 50-341 Replace the following pages of the Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by Amendment number and contain a vertical line indicating the area of change.

REMOVE INSERT 8 2-7 0 2-7 3/4 3-1 3/4 3-1 3/4 3-la 3/4 3-2 3/4 3-2 3/4 3-5 3/4 3 5 3/4 3 6 3/4 3-6 3/4 3-7 3/4 3-7 3/4 3-8 3/4 3 8 3/4 3-9 3/4 3 9 3/4 3-11 3/4 3-11 3/4 3-13 3/4 3-13 3/4 3-14 3/4 S-14 3/4 3-14a 3/4 3-14a 3/4 3-20 3/4 3-20 3/4 3-21 3/4 3-21 3/4 3-22 3/4 3-22 3/4 3-25 3/4 3-i5 3/4 3-26 3/4 3-26 3/4 3-30 3/4 3-30 3/4 3-31 3/4 3-31 3/4 3-37 3/4 3-37 3/4 3-38 3/4 3-38 i

3/4 3-39*

3/4 3-40 3/4 3-40 3/4 3-45 3/4 3-45 1

Overleaf page provided to maintain document completeness.

No changes contained on this page.

UMITING SAFETY SYSTEM SETTINM BASES REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued)

Average Power Ranae Monitor (Continued)

Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks and because several rods must be moved to change power by a significant amount, the rate of power rise is very slow.

Generally the heat flux is in near equilibrium with the fission rate.

In an assumed uniform rod withdrawal approach to the trip level, the rate of power rise is not more than 5% of RATED THERMAL POWER per minute and the APRM system would be more than 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 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 system and therefore the monitors respond directly and quickly to changes due to transient operation for the case of the fixed Neutron flux Upscale setpoint; i.e. for a power increase, the THERMAL POWER of the fuel will be less than that indicated by the neutron flux due to the time constants of the heat transfer associated with the fuel, for the flow Biased Simulated Thermal Power - High setpoint, a time constant of 6 i ! seconds is introduced into the flow biased APRM in order to simulate the fuel therit.al transient characteristics.

A more conservative maximum value is used for the flow biased setpoint as shown in Table 2.2.1-1.

The APRM setpoints were selected to provide adequate margin for the Safety Limits and yet allow operating margin that reduces the possibility of unnecessary shutdown.

for single recirculation loop operation, the reduced APRM setpoints are based on a A W value of 8%.

The A W value corrects for the difference in indicated drive flow (in percentage of drive flow which produces rated core flow) between two loop and single loop operation of the same core flow.

The decrease in setpoint is derived by multiplying the slope of the setpoint curve by 8%. The High flow Clamped flow Biased Neutron flux-High setpoint is not applicable to single loop operation as core power levels which would receire this limit are not achievable in a single loop configuration.

l 3.

Reactor V11sel Steam Dome Pressure Hiah High pressure in the ' nuclear system could cause a rupture to the nuclear system process barrier resulting in the release of fission products.

A pressure increase while operating will also tend to increase the power of the reactor by compressing voids thus adding reactivity.

The trip will quickly reduce the neutron flux, counteracting the pressure increase.

The trip setting is slightly higher than the operating pressure to permit normal operation without spurious FERhl UNIT 2 B 2-7 Amendment Ho, p, Q,75

3/4.3 INSTRUBENTA1103 3/4.3.1 RE ACTOR PROT ECILORJ1EU3.JN11RUME NT AT 10N llHlTING CONDITION FOR OPERATION 3.3.1 As a minimum, the reactor protection system instrumentation chainels shown in Table 3.3.1-1 shall be OPERABLE with the REAC10R PROTECTION SYSllM RESPONSE TIME as shown in Table 3.3.1 2.

APPLICABIL111:

As shown in Table 3.3.1-1.

ACTION:

a.

With the number of OPERABLE channels less than required by the Minimum OPERABLE channels per Trip System requirement for one trip system:

1.

Within I hour, verify that each functional Unit within the affected trip system contains no more than one inoperable channel or place the inoperable channel (s) and/or that trip system in the tripped condition'.

2.

If placing the inoperable channel (s) in the tripped condition would cause a scram, the inoperable channel (s) shall oc restored to OPERABLE status within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or the ACTION required by Table 3.3.1-1 for the affected functional Unit shall be taken.

3.

If placing the inoperable channel (s) in the tripped condition would not cause a scram, place the inoperable channel (s) and/or that trip system in the tripped condition within 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.

b.

With the number of OPERABLE channels less than required by the liinimum OPERABLE Channels per Trip System requirement for both trip systems, place at least one trip system ** in the tripped condition within I hour and take the ACTION required by Table 3.3.1-1.

An inoperable channel need not be plar.ed in the tripped conditior, where this would cause a scram to occur.

In these cases, the inoperable channel shall be restored to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after the channel was first determined to be inoperable or the ACTION required by Table 3.3.1-1 for that functional Unit shall be taken.

- The trip system need not be placed in the tripped condition if this would cause a scram to occur.

When a trip system can be placed in the tripped condition without causing a scram to occur, place the trip system with the most inoperable charrfis in the tripped condition; if both systems have the same number of inoperabie channels, place either trip system in the tripped condition.

FERH1 - UNIT 2 3/a 3-1 Amendment No.75

~-

l 3/4.3 INSTRUMENTATIQS 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SVRVElllANCE RE0VIREMENTS 4.3.1.1 Each reactor 3rotection system instrumentation channel shall be demonstrated OPERABLE yy the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.1.1 1.

4.3.1.2 LOGIC SYSTEM FUNCTIONAL TLSTS 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 functional unit shown in Table 3.3.12 shall be demonstrated to be within its limit at least once per 18 months.

Each test shall include at least one channel per trip system such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific reactor trip system.

j l

t

(

l FERMI - UNIT 2 3/4 3-la Amendment No. 75 l i

TABLE 3.3.1-1 I

REACTOR PROTECTION SYSTEM INSTRUMENTATION i

APPLICABLE MINIMUM

,,g-OPERATIONAL OPERABLE CHANNE S) 3 FUNCTIONf,l UNIT CONDITIONS PER TRIP SYSTEM 8 ACTION h

1.

Intermediate Range Monitors (D)-

5 a.

Neutron Flux - High 2

3 1

3, 4 5(C) 3(d) 2 3

3 i

b.

Inoperative 2

3 1

t 3, 4 3

2 l

5 3(d) 3 l

2.

Average Power Range Monitor (e):

1 a.

Neutron fiux - high, Setdown 2

2 1

3(C) 2(d) 3 L

2 2

l 5

w b.

Flow Biased Simulated Thermal 1

2 4

l 1

Power - High I

c.

Fixeo Neutron Flux - High 1

2 4

d.

Inoperative 1, 2 2

1 3

2 2.

5(C) 2(d) 3 i

3.

Reactor Vessel Steam Dome l

Pressure - High 1, 2(f) 2 1

4.

Reactor Vessel low Water Level -

Level 3 1, 2 2

1 i

g 5.

Main Steam Line Isolation v'alve -

4 g

Closure 1(9) 4 4

E n

(11 i

f l

MOLE 3.3.1 1 (Continued)

REACTOP PR01ECT10N SYSTEM INSTRUMENTAllM MBLE NOTATIE S (a) A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for j

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) This function shall be automatically bypassed when the reactor mode switch is in the Run position.

(c) Unless adequate shutdown margin has been demonstrated per Specification 3.1.1, the " shorting links" shall be removed from the RPS circuitry prior to and during the time any control rod is withdrawn.*

(d) When the ' shorting links

are removed, the Minimum OPERABLE Channels Per Trip System is 4 APRMs, 6 1RMs and per Specification 3.9.2, 2 SRMs.

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

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

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

(h) This function is not required to be OPERABLE when PRIMARY CONTAINMENT INTEGRITY is not required.

(i) With any control rod withdrawn. Not applicable to control rods removed per Specificaticn 3.9.10.1 or 3.9.10.2.

(d) This function shall be automatically bypassed when turbine first stage pressure is < 154 psig, equivalent to THERMAL POWER less than 30% of RATED THERMA [ POWER.

Not required for control rods removed per Specification 3.9.10.1 or 3.9.10.2.

FERMI - UNIT 2 3/4 3-5 Amendment No. 75

TABLE 3.3.1-2 l

REACTOR PROTECTION SYSTEM RESPONSE TIMES 9

' 3 RESPONSE TIME FUNCTIONAL UNIT..

(Seconds) i c5

1.. Intemediate Range Monitors:

a. -

Neutron' Flux - High NA

~

NA' b.

Inoperative 2.

Average Power Range Monitor *:

a a.

Neutron Flux'- High, Setdown MA b.

Flow Biased Simulated Thermal Power - High 6 i 1**

I

'l c.

Fixed Neutron Flux - High

< 0.09 d.

Inoperative NA

'3.

Reactor Vessel Steam Dome Pressure - High

< 0.55 j

4.

Reactor Vessel Low Water Level - Level 3 7 1.05 5.

Main Steam Line Isolation Valve - Closure 7-0.06 R

6.

Main Steam Line' Radiation - High.

NA 7.

Drywell Pressure - High NA Y

8.

Scram Discharge Volume Water Level - High a.

Float Switch NA b.

Level Transmitter NA 9.

Turbine Stop Valve - Closure

< 0.06

10. Turbine Control Valve' Fact Closure 7 0.08***

11. Reactor Mode Switch Shutcown Position HA 12.. Manual Scram NA' y

13. Deleted' E

i i

a l

l

Neutron detectors are exempt from response time testing. Response time shall be measured j!'

- Including simulated ti,ermal power time constant.

from the detector output or from the input of the first electronic component in the channel..

I D

- - Measured from deenergization of K-37. relay which inputs the turbine control valve closure signal y

to the RPS.

m l

t i

.--3

TABLE 4.3.1.1-1 4

L-REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS g

CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH l

[

FUNCTIONAL UNIT CHECK TEST CALIBRATION (a)

SURVEILLANCE REQUIRED 1.

. Intermediate Ranga Monitors:

a.

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

S/U(C),W SA 2

i S

W SA 3, 4, S i

b.

Inoperative NA W

NA 2,3,4,S

[

r 2.

Average Power Range Monitor (f):

a.

Neutron Flux -

S/U,S,(b)

S/U(c),W SA 2

High, Setdown S

W SA 3, S R

b.

Flow Biased Simulated Thermal Power - High S

S/U(C), Q W(d)(e),SA,R(h) j Y

c.

Fixed Neutron Flux -

High S

S/U(C), Q W(d),SA 1

d.

Incperative NA Q

NA 1, 2, 3, 5 I

3.

Reactor Vessel Steam Dome Pressure - High S

Q(k)

R 1, 2 4.

Reactor Vessel Low Water Level - Level 3 S

Q(k)

R 1, 2

&g 5.

Main Steam Line Isolation Valve - Closure NA Q

R 1

E 6.

Main Steam Line Radiation -

p High S

Q R

I,2(1)

M 7.

Drywell Pressure - High 5

0(k)

R 1, 2 i

l TABLE 4.3.1.1-1 (Continued) m

o3 REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE PE0UIREMENTS C3 CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH FUNCTIONAL UNIT CHECK TEST CALIBRATION SURVEILLANCE REQUIRED 8.

Scram Discharge Volume Water Level - High a.

Float Switch NA Q@)

R 1, 2, 5(j) b.

Level Transmitter S

Q R

1, 2, 5(j) 9.

Turbine Stop Valve - Closure NA Q

R

}

10. Turbine Control Valve Fast Closure NA Q

NA 1

R

11. Reactor Mode Switch Shutdown Position NA R

NA 1,2,3,4,5 co

12. Manual Scram NA W

NA I,2,3,4,5 S

13. Deleted.

E.

M (a)

Neutron detectors may be excluded from CHAN'sEL CALIBEATIO's.

5 (b)

The IRM and SRM channels shall b= detemined to overlap for at least 4 & cades during each startuo af ter enttring CPERATIC4AL CCCITIO't 2 a,d the IR" and APRM channels shall be detemined to overlag for at least % deca &s during each centro 11ed shutdow,. If not perfomed with% the previous 7 days.

,^o (c)

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to startup, if rot perforaM within the previous 7 <tays..

(d)

This calibration shall constst of the adjustment of the AFFM channel to etnform to the pe=er values calevlated t-a Nat balance ouete g CPERATIO4AL w.

C04DITIO4 I when THER*AL POWER r,25% of RATED THER% KNER. Adtust the #PEM channel if the absolute dif ference is greater than 2% cf SATED THE8"At

?

POWER.

(e)

This calibration shall consist of the adjustwnt of the APR* ??i.;.- SW chem el to conform to a cat tbrated flow signal.

D (f)

The LPRMs shall be calibrated at least once per 1000 effective full pe=er hours (EFPH) usteg the TIP system.

(g)

Deleted.

w (h)

This calibration shall consist of vertfying the 6 + 1 second simulated tNmal power tiae constant.

7 (t)

This function is not required to be O'4RABLE when the reacter pressure vt3sel head is re-eved per Specificattos 3.10.1.

y (j)

With any control rod withdrawn. not appitcable to control rods remed per Spact fication 3.9.10.1 er 3.9.10.2.

(k)

Includes vertitcation of the trip setpoint of the trip unit.

l

IESTRUMENTATION 3/4.3.2 ISOLATION AEJUATION INSTRUMENTATION tiMITING CONDITION FOR OPERATION 3.3.2 The isolation actuation instrumentation channels shown in Table 3.3.2-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.2-2 and with ISOLATION SYSTEM RESPONSE TlHE as shown in Table 3.3.2-3.

APPLICABILITY:

As shown in Table 3.3.2-1.

A(llati:

a.

With an isolation actuation instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.2-2 declare the channel inoperable until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.

b.

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for one trip system:

1.

If placing the inoperable channel (s) in the tripped condition would cause an isolation, the inoperable channel (s) shall be restored to OPERABLE status within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or the ACTION required by Table 3.3.2-1 for the affected trip function shall be taken.

2.

If placing the inoperable channel (s) in the tripped condition would not cause an isolation, the inoperable channel (s) and/or that trip system shall be placed in the tripped condition within:

a) 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for trip functions common to RPS Instrumentation; and b) 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for trip functions not common to RPS Instrumentation.

The provisions of Specification 3.0.4 are not applicable, c.

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 j

within one hour and take the ACTION required by Table 3.3.2-1.

Place one trip system (with the most inoperable channels) in the tripped condition.

The trip system need not be placed in the tripped condition when this would cause the isolation to occur.

FERMI - UNIT 2 3/4 3-9 Amendment No. /J,75

TABLE 3.2.2-1 r,

E ISOLATION ACTUATION INSTRUMENTATION VALVE GROUPS MINIMUM APPLICABLE E

OPERATED BY OPERABLE CHANNE S)

OPERATIONAL Q

TRIP FUNCTION SIGNAL PER TRIP SYSTEM a CONDITION ACTION m

1.

PRIMARY CONTAINMENT ISOLATION Reactor Vessel L w# Water Level a.

1)

Level 3 (e '

13, 15 2

1, 2, 3 20 l

2)

Level 2 (d) 2, 12, 14, 16, 17, 18 2

1, 2, 3 20 3)

Level I I

2 1, 2, 3 20 b.

Drywell Pressure - High (3)## 2, 12, 13, 14, 15, 16, l

17, 18 2

1, 2, 3 20 w

c.

Main Steam Line x

1)

Radiation - High 1, 2 2

1, 2, 3 21 l

w J

2)

Pressure - Low 1

2 1

22 3)

Flow - High 1

2 1, 2, 3 21

~

d.

Main Steam Line Tunnel Temperature - High 1

2(C)

I, 2, 3 21 e.

Condenser Pressure - High 1

2 1,

2**, 3**

21 f.

Turbine Bldg. Area y

Temperature - High 1

2 1, 2, 3 21 3

g.

Deleted E

h.

Manual Initiation 1, 2, 3, 5, 12, 13, 14 1/ valve 1, 2, 3 26 j

g 15, 16, 17. 18 4

i-TABLE 3.3.2-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION E

VALVE GROUPS MINIMUM APPLICABLE f

OPERAltD BY OPERABLE CHANNE S)

OPERATIONAL i

TRIP FUNCTION SIGNAL PER TRIP SYSTEM i CONDITION _

ACTION c.5 4.

HIGH PRESSURE COOLANT INJECTION SYSTEM ISOLATION a.

HPCI Steam Line Flow - High 1.

Differential Pressure 6

1 1, 2, 3 23 2.

Time Delay 6

1 1,2,3 23 i

b.

HPCI Steam Supply Pressere - Low 6, 7 (9) 2 I, 2, 3 23 j

c.

HPCI Turbine Exhaust Diaphragm Pressure - High 6

2 1, 2, 3 23 d.

HPCI Equipment Room Tenperature - High 6

I 1, 2, 3 23 e.

Manual Initiation 6, 7 I/ valve I, 2, 3 26 5

S.

RHR SYSTEM SHUTD0hw COOLING MODE ISCL*, TION i

Reactor Vessel Low # Water a.

Level - Level 3' 4Ie) 2 1, 2, 3 25 I

b.

Reactor Vessel (Shutdown Cooling Cut-in Permissive Interlock)

Pressure - High

?

i 1, 2, 3 25 c.

Manual Initiation 4

1/ valve 1, 2, 3 26

$"g 6.

SECONDARY CONTAINMENT ISCLATION a.

ReactorVesselLgy)Wger 9) level - Level 2 (D g

(

2 1, 2, 3, and

24 y

b.

Drywell Pressure - High (b) (j)J8 2**

2 1, 2, 3 24 i

D c.

Fuel Pool Ventilation Exhaust)

Radiation - High (b m

2 1, 2, 3, and

24 d.

Manual Initiation (b) 1(i)

I, 2, 3, and

27 3

y e

- + -

w-

JABLE 3.3.2-1 (Continued)

ISOLATION AC10AT10N INSTRUMENTATION ACTION STATEMENTS Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD ShU1DOWN ACTION 20 within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Be in at least STARTUP with the associated isolation valves closed ACTION 21 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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

ACTION 22 Close the affected system isolation valves within I hour and ACTION 23 declare the affected system inoperable.

ACTION 24 Establish SECONDARY CONTAINMENT INTEGRITY with the standby gas i

treatment system operating within I hour.

ACTION 25 Disable in the closed position the affected system isolation valves within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and declare the shutdown cooling mode of RHR inoperable.

Restore the manual initiation function to OPERABLE status within 8 ACTION 26 hours3.009259e-4 days 0.00722 hours 4.298942e-5 weeks 9.893e-6 months or close the affected system isolation valves within the next hour and declare the affected system inoperable.

t ACTION 27 Restore the manual initiation function to OPERABLL status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or establish SECONDARY CONTAINMENT INTEGRITY with the Standby Gas Treatment System operating.

TABLE NOTATIONS When handling irradiated fuel in the secondary containment, during CORE ALTERATIONS, or during opet ations with a potential for draining the reactor vessel.

The high condenser pressure input to the isolation actuation in trumentation may be bypassed during reactor shutdown or for reactor startup when condenser pressure is above the trip setpoint.

Actuates dampers shown in Table 3.6.5.2 1.

(a)

A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for l

required surveillence without placing the channel or tri) system in the tripped condition provided at least one other OPERABLE ciannel in the same trip system is monitoring that parameter.

In addition, for the HPCI system and RCIC system isolation, provided that the redundant isolation valve, inboard or outboard, as applicable, in each line is uPERABLE and all required actuation instrumentation for that valve is OPERABLE, one channel may be placed in an inoperable status for up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for required surveillance without placing the channel or trip system in the tripped ccndition.

(b)

Also starts the standby gas treatment system.

L (c)

A channel is OPERABLE if 2 of 4 detectors in that channel are OPC'1ABLE.

L (d) level signal actuates Groups 2, 10, 11, 12, 14, 16, 17, 18, and (e)

This level signal actuates Groups 4,13 and 15.

FERMI:- V'IT 2 3/4 5-14 Amendment No. fi,75 N

BilLL3.3.21 (Continued) 110MT10N AC1UAT10N lfiSIRUMENTAllDti 1ABLENOTATIONS(Continued)

(f)

IsolateswithsimultaneousRCICSteamSupplyPressureLow(Isolation instrumentation and Drywell Pressure High (ECCS Actuation Instrumentation.

(9)

Isolates with simultaneous HPCI Steam Supply Pressure tow (1 solation 1

Actuation Instrumentation) and Drywell Pressure High (ECCS Actuation j

instrumentation).

(h) Reserved.

(1)

Secondary Containment Isolation Push buttons.

(J) h s pressure signal actuates Groups 2, 12, 13, 14, 15, 16, 17, 18, and With time delay of 45 seconds.

8 These trip function (s) are common to the RPS trip function.

l s

FERMI - UNIT 2 3/4 3-14a Amendment No. /J,75

TABLE 4.3.2.1-1 b

ISOLATION ACTUATION INSTRtMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPEPATIONAL 25 CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH e

[

TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REOGIRED 1.

PRIMARY CONTAINMENT ISOLATION Reactor Vessel Low Water Level-a.

1)

Level 3 5

Q R

I, 2, 3 i

R 1, 2, 3 2)

Level 2 5

Q' 3)

Level I S

Q R

1, 2, 3 b.

Drywell Pressure - High S

Q' R

1, 2, 3 c.

Main Steam Line Y

1)

Radiation - High 5

Q R

1, 2, 3 2)

Pressure - Low 5

Q R

I 3)

Flow - High 5

Q' R

1, 2, 3 o

d.

Main Steam Line Tunnel Temperature - High S

Q R

1, 2, 3 e.

Condenser Pressure - High 5

0 R

1, 2 * *, 3 *

8 f.

Turbine Bldg. Area l

Temperature - High 5

Q R

1, 2, 3 8

ct j

g.

Deleted z

h.

Manual Initiation NA R

NA 1, 2, 3 n

h

e TABLE 4 3.2.I-I (Continued)

ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS 9

CHANNEL OPERATIONAL

?i CHANNEL FUNCTI0ftAL CHANNEL CONDITIONS FOR WHICH TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED 2.

REACTOR WATER CLEANUP SYSTEM ISOLATION a.

A Flow - High 5

Q R

1, 2, 3 b.

Heat Exchanger / Pump /High Energy Piping Area Temperature - High S

Q R

1, 2, 3 c.

Heat Exchanger / Pump / Phase Separator Area Ventilation A Temperature - High S

Q R

1, 2, 3 d.

SLCS Initiation NA R

NA 1, 2, 3 M

^

e.

Reactor Vessel L.,.: Water 8

Level - Level 2 5

Q R

1, 2, 3 l

f.

Deleted 9

Manual Initiation NA R

NA 1, 2, 3 3.

REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION k

a.

RCIC Steam Line Flow - High E

1.

Differential Pressure S

Q' R

1, 2, 3 2

2.

Time Delay NA Q

R 1, 2, 3 m

b.

RCIC Steam Supply Pressure -

P Low 5

Q R

I, 2, 3 c.

RCIC Turbine Exhaust Diaphragm g

Pressure - High S

Q R

1, 2, 3 O

d.

RCIC Equipment Room 8

Temperature - High S

0 R

1, 2, 3 e.

Manual Initiation NA R

NA I, 2, 3 i

~

i TABL E 4.3.2.i.' (Continued)

g ISOLATION ACTUATION INSTPDMENTATilN SURVEILLANCE REQUIREMENTS 3E

~~

CIANNEL OPERATIONAL CHANNEL FUNC~i '<NAL CHANNEL CONDITIONS FOR WHICH ji TRIP FUNCTION CH2CK TEST CALIBRATION SURVEllLANCE REQUIRED C

4.

HIGH PRESSURE COOLANT INJECTION SYSTEM ISOLATION n,

a.

HPCI Steam Line Flow - High 1.

Differential Pressure S

Q' R

1, 2, 3 2.

Time Delay NA Q

R 1, 2, 3 b.

HPCI Steam Supply Pressure - Low S

Q R

1, 2, 3 c.

HPCI Turbine Exhaust Diaphragm Pressure - High S

Q,'

R 1, 2, 3 d.

HPCI Equipment Room Temperature - High 5

Q' R

1, 2, 3 e.

Manual Initiation NA R

NA 1, 2, 3 u,h 5.

RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION u,

4, a.

Reactor Vessel Low Water Level -

Level 3 5

Q R

1, 2, 3 b.

Reactor Vessel (Shutdown Cooling Cut-in Permissi'e Interlock)

Pressure - High S

Q R

1, 2, 3 c.

Manual Initiation NA R

NA 1, 2, 3 6.

SECONDARY CONTAINMENT ISOLATION a.

Reactor Vessel Low Water Level -

gr Level 2 5

Q R

1, 2, 3, and

g b.

Drywell Pressure - High S

Q R

1, 2, 3 jt c.

Fuel Pool Ventilation Exhaust g

Radiation - High S

Q R

1, 2, 3, and

d.

Manual Initiation NA R

NA 1, 2, 3, and

c*

5 2

When handling irradiated fuel in the secondary containment, during CORE ALTERATIONS, and during operations with a potential for draining the reactor vessel.

y

- May be bypassed under administrative control.

Includes verification of the trip setpoint of the trip unit.

l

r TABLE 3.3.3-1 (Continued)

EMERGENCY CORE COOLING SYSTEM ACTUATf04 INSTRLHENTATION C5 MINIMUM GPERABLE CHANNELS PER APPLICABLE

~

TRIP OPERATIONAL SYSTEM (a)

CONDITIONS ACTION TRIP FUNCTION 4.

AUTOMATIC DEPrtESSURIZATION SYSTEM a.

Reactor Vessel Low Water Level - Level 1 2

1, 2, 3 30 2

1, 2, 3 30 b.

Drywell Pressure - High I

1, 2, 3 31 c.

ADS Timer d.

Core Spray Pump Discharge Pressure - High (Permissive) 1/ pump 1, 2, 3 31 RHR LPCI Mode Pu p Discharge Pressure - High e.

(Pemissive)

I/pu=p 1, 2, 3 3I

)

f.

Reactor Vessel Low Water Level - Level 3 (Permissive) 1 1, 2, 3 31 g.

Manual Initiation I/ valve I, 2, 3 33 h.

Drywell Pressure - High Bypass Timer 2

1, 2, 3 31 I

1, 2, 3 33 g,

m i.

Manual Ir.hibit MINIMUM APPLICABLE TOTAL NO.

CHANNELS CHANNELS OPERATIONAL OF CHAMELS TO TRIP OPERABLE CONDITIONS ACTION 5.

LOSS OF POWER y

1.

4.16 kV Emergency Bus Under-g voltage (Loss o' Voltage) 2/ bus I/ bus I/ bus I, 2, 3, 4**, 5**

35 2.

4.16 kV Emergercy Bus Under-g.

g voltage (Degraded Voltage) 2/ bus 1/ bus I/ bus I, 2, 3, 4**, 5**

35 A charmel may be p'eced in en irepareble status for be to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for required surveillance without placing the trip system in t*e tripk*d l

I ja) condition providad at least ore CPERABLE chamel in the same trip syste is wc94toring that para =*ter.

2o (b)

Also actuates the associated emettymcy diesel g*-eraters, g

(c)

One trip system. Providas signals to HPCI afd RCIC suction valves.

(d)

One trip systee. Provides signal to Mcl ptro suctic9 valves only.

.y (e)

On 2 out of 2 logic, provides a signal to trip the HPCI turbice.

When the syst-m is required to be CPERABLE per Specification 3.s.2.

Required when EsF equipr= nt is required to be OPEcASLE.

Not required to be OPE 8ABLE when reacter steam dome r ssun is less then or eat to 153 pstv.

ff imftvidual ca w t controls.

B BLE 3.3 E (Continued)

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMffilA11M KTION STATEMENTS ACTION 30 -

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Systen requirement:

a.

For one trip system, place that trip system in the tripped condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

or declare the associated ECCS l

inoperable, b.

For both trip systems, declare the associated ECCS inoperable.

ACTION 31 -

With the number of CPERABLE chtonels less than required by the Minimum OPERABLE Channels per Trip System requirement, declare the associated ADS Trip System inoperable within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l ACTION 32 -

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip system requirement, place the inoperable channel in the tripped condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 33 -

Restore the manual initiattun and/or manual inhibit function to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or declare the associated ECCS l

or ADS Trip System inoperable.

ACTION 34 -

With the number of OPERABLE chanrt'lt less than required by the Minimum OPERABLE Channels per Trip System requirement, place at least one inoperable channel in the tripped condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months *, align the HPCI system to take suction from the j

suppression pool, or declare the HPCI system inoperable.

ACTION 35 -

With the number of OPERABLE channels:

a.

One less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or declare the associated emergency diesel generator inoperable and take the ACTION required by Specification 3.8.1.1 or 3.8.1.2, as appropriate, b.

Less than the Minimum Channels OPERABLE requirement, declare the associated diesei generator inoperable and take the ACTION required by Specification 3.8.1.1 or 3.8.1.2, as appropriate.

The provisions of Specification 3.0.4 arc not applicable.

FERMI - UNIT 2 3/4 3 26 Amendment No. U, 75 l

VABLE 4.3.3.X-1 e

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIRE *'ENTS CHANNEL OPERATIONAL i

9 CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH 3

TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED 1.

CORE SPRAY SYSTEM c

a.

Reactor Vessel Low Water Level -

5 Level 1 5

Q R

1, 2, 3, 4*, 5*

t l

[

b.

Drywell Pressure - High 5

Q R

1, 2, 3 I

c.

Reactor Steam Dome Pressure -

Low S

Q*#

R 1, 2, 3, 4*, 5*

d.

Manual Initiation NA R

NA 1, 2, 3, 4*, S*

2.

LOW PRESSURE COOLANT INJECTION MODE OF RHR SYSTEM a.

Reactor Vessel Low Water Level -

0 "'

R 1, 2, 3, 4*, 5*

Level'l S

b.

Drywell Pressure - High S

Q' R

1, 2, 3 c.

Reactor Steam Dome Pressure -

Low 5

Q'#

R 1, 2, 3, 4*, 5*

d.

Reactor Vessel Low Water Level - Level 2 S

Q R

1, 2, 3, 4*, 5*

e.

Reactor Steam Dome Pressure -

[

Low S

Q

R 1, 2, 3, 4*, 5*

o f.

Riser Differential Pressure -

High S

Q R

1, 2, 3 g.

Recirculation Pump Differential Pressure - High S

Q*#

R 1, 2, 3 h.

Manual Initiation NA R

NA 1, 2, 3, 4*, 5*

3.

HIGH PRESSURE COOLANT INJECTION SYSTEM' Raactor Vessel Low Water Level -

a.

Q#

R 1, 2, 3 Level 2 S

b.

Drywell Pressure - High S

Q R

I,2,3 c.

Condensate Storage Tank Level -

Low C

Q" R

1, 2, 3 fg d.

Supprusion Pool Water Level -

g High S

Q R

1. 2, 3 3

e.

Reactor Vessel High Water Level -

Levei 8 5

Q

R 1, 2, 3

.o f.

Manual Initiation HA R

NA 1, 2, 3 s

e TABLE 4.3.3.1-1 (Continued)

A EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE FEDUIREMENTS E

CHANNEL OPERATIONAL

~

CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH E

TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE PEOUIFED

~w 4.

AUTOMATIC DEPRESSURIZATION SYSTEM

~

a.

Reactor Vessel Low Water Level -

8#

Level 1 S

Q R

1, 2, 3 b.

Drywell Pressure - High 5

Q R

I, 2, 3 c.

ADS Timer NA Q

R 1, 2, 3 d.

Core Spray Pump Discharge Pressure - High S

Q'#

R 1, 2, 3 e.

RHR LPCI Mode Pump Discharge Pressure - High 5

Q" R

1, 2, 3 f:

Reactor Vessel Low Water Level -

8#

Level 3 S

Q R

1, 2, 3 w

g.

Hanual Initiation NA R

NA I, 2, 3 1

h.

Drywell Pressure - High Bypass

~imer NA 0

R 1, 2, 3 1

r 4,

i. Manual Inhibit NA R

NA I, 2, 3 w

5.

jfSS Or POWER a.

4.16 kV Eme=gency Bus Under-voltage (Loss of Voltage)

(Division 1 and Division 2)

NA M

R I, 2, 3, 4**, 5**

b, 4.16 kV Emergency Bus Under-F voltage (Degraded Voltage) 8 (Division 1 and Division 2)

NA M

R I, 2, 3, 4**, 5**

a When the syster, is required to be OPERABLE per Specification 3.5.2.

7 Requimd OPERABLE when ESF equipment is required to be OPERABLE.

Not required te be OPERABLE when reactor steam dome pressure is less than or equal to 150 psig.

u l

v, Include verification of the trip setpoint of the trip unit.

i j

a A

E 7

TABLE 3.3.5-1 E

REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION Z

MINIMUM m-OPERABLE CHANNELS )

FUNCTIONAL Ut(ITS PER TRIP SYSTEMga ACTION a.

Reactor Vessel Low Water Level - Level 2 2

50 b.

Reactor Vessel High Water Level - Level 8 2(b) 50 c.

Condensate Storage Tank Water Level - Lew 2(C) 51 d.

Manual Initiation 1/ valve 52 R.

to (a) A channel cay be placed in an inoperable status for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for required surveillance without l

placing the trip system in the tripped condition provided at least one other OPERABLE channel in the same trip system is monitormg that parameter.

(b) One trip system with two-out-of two logic.

(c) One trip system with one-out-of-two logic.

E R

a n

b e

TABLE 3.3.5-1 (Continued)

REACTOR CORE ISOLATION COOLING SYSTEM ACTION STATEMENTS ACTION 50 -

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement:

a.

For one trip system, place the inoperable channel (s) and/or that trip system in the tripped condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or declare the RCIC system inoperable, j

b.

For both trip systems, declare the RCIC system inoperable.

ACTION 51 -

With the number of OPERABLE channels less than required by the Minimum OPEPABLE Channels per Trip System requirement, place at least one inoperable channel in the tripped condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

or align RCIC to take suction from the suppression l

pool or declare the RCIC system inoperable.

ACTION 52 -

Restore the manual initiation function to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or declare the RCIC system inoperable.

l

The provisions of Specification 3.0.4 are not applicable.

FERMI - UNIT 2 3/4 3-38 Amendment No. 75

TABLE 3.3.5-2

?

REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS ALLOWABLE E

FUNCTIONAL UNITS TRIP SETPOINT VALUE N

a.

Reactor Vessel Low Water Level - Level 2 1 110.8 inches *

> 103.8 in-hes b.

Reactor Vessel High Water Level - Level 8 5 214 inches

i 219 inches c.

Condensate Storage Tank Level - Low 2 3 inches 1 0 inches (27 inches above (24 inches above tank bottom) tank bottom) d.

Manual Initiation NA NA T

s T

See Bases Figure B 3/4 3-1.

9

P

_ TABLE 4.3.5.1-1 REACTOR CORE ISutATION COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS E-CHANNEL CHANNEL FUNCTIONAL CHANNEL FUNCTIONAL UNITS CHECK TEST CALIBRATION a.

Reactor Vessel Low Water Level - Level 2 S

Q R

I b.

Reactor Vessel High Water S

Q R

Level - Level 8 c.

Condensate Storage Tank level - Low S

Q R

,s*

d.

Manual Initiation NA R

NA k

Includes verification of the trip setpoint of the trip unit.

E a

a n

s TABLE 4.3.6-1 CONTROL R0D BLOCK INSTRUMENTATION SURVEILLANCE RE0UIREMENTS 93 CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH TRIP FUNCTION CHECK TEST CALIBRATION (a)

SURVEILLANCE REQUIRED c-22i 1.

ROD BLOCK MONITOR a.

Upscale NA S/U Q

Q 1*

S/U(D,Q b.

Inoperative NA

~

NA 1*

S/U b),q q

1*

c.

Downscale NA 2.

APRM a.

Flow Biased Neutron Flux -

High S

S/U(b),Q 5A I

b.

Inoperative NA S/U(b),Q NA 1, 2, 5 c.

Downscale S

S/U(b),g sg j

d.

Neutron Flux - Upscale, Setdown 5

S/U(b),Q SA 2, 5 3.

SOURCE RANGE MONITORS S/U(b)

NA 2***,

5 R

a.

Detector not full in NA S/U(b),W b.

Upscale S

S/U(b),,W SA 2***,

5 Y

c.

Inoperative NA W

NA 2***,

5 C,

d.

Downscale S

S/U(b),W SA 2***,

5 4.

INTERMEDIATE RANGE MONITORS a.

Detector not full in NA S/U(b),W NA 2, 5 b.

Upscale S

S/U(b),W SA 2, 5 c.

Inoperative NA S/U(b),W NA 2, 5 d.

Downscale S

S/U(b),W SA 2, 5 5.

SCRAM DISCHARGE VOLUME a.

Water Level - High NA.

Q R

I, 2, 5**

b.

Scram Trip Bypass NA R

NA 2, 5**

3 6.

REACTOR COOLANT SYSTEM RECIRCULATION FLOW S/U((b),g a.

Upscale NA g

j S/U b)

NA 4

3 b.

Inoperative NA S/U(b),Q c.

Comparator NA

,g g

j 7.

REACTOR MODE SWITCH y

SHUTOOWN POSITION NA R

NA 3, 4 m

Person / Time
ML20082U403
Site:	Fermi (NPF-43-A-075, NPF-43-A-75)
Issue date:	09/06/1991
From:	Marsh L Office of Nuclear Reactor Regulation
To:
Shared Package
ML20082U406	List: ML20082U403 ML20082U427
References
NUDOCS 9109200152
Download: ML20082U403 (28)
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ML20082U403

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