ML19347A757

Proposed Tech Specs 1.0,3.3 & 4.3 Re Reactor Protection Sys

ML19347A757
Person / Time
Site:	Browns Ferry
Issue date:	09/24/1980
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML19347A752	List: ML19347A751 ML19347A757
References
NUDOCS 8009300156
Download: ML19347A757 (15)

Text

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j ENCLOSURE 1 PROPOSED CHANGES TO TECHNICAL SPECIFICATIONS

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BROWNS FERRY NUCLEAR PLANT (DOCKET NOS.' 50-259, -260, -296) 1 i

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O UNITS 1 AND 2 PROPOSED CHANGES I

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1.0 DEFINITIONS (Cont'd) 2.

Run Mode - In this mode the reactor system pressure is at or above 825 pa and the reactor protection system is energized with APRM protection (excluding the 155 high flux trip) and RBM interlocks in service.

3.

Shutdown Mode - Placing the mode switch to the shutdown position initiates a reactor scram and power to the control rod drives is removed. After a short time period (about 10 sec), the scram signal is removed allowing a scram reset and restoring the normal valve lineup in the control rod drive hydraulic system; also, the main steam line isolation scram and main condenser low vacuum scram are bypassed if reactor vessel pressure is below 1055 psig.

4.

Refuel Mode - With tne mode switch in the refuel position 2

interlocks are established so that one control rod only may be withdrawn when the Source Range Monitor indicate at least 3 cps and the refueling crane is not over the reactor; also, the main steam line isolation scram and main condenser low vacuum scram are bypassed if reactor vessel pressure is below 1055 psig. If the refueling crane is over the reactor, all rods must be fully inserted ard none can be withdrawn.

N.

Rated Power - Rated power refers to operation at a reactor power of 3,293 MWt; this is also termed 100 percent power and is the maximum power level authorized by the operating lisense. Rated steam flow, rated coolant flow, rated neutron flux, and rated nuclear system pressure refer to the values of these parameters wh<.n the reactor is at rated power. Design power, the power to whien the safety analysis applies, corresponds to 3,440 Mwt.

1 0.

Primary Containment Integrity - Primary containment integrity means that the drywell and pressure suppression chamber are f.ntact and all i

of the following conditions are satisfied:

1.

All non-automatic containment isolation valves on lines connected to the reactor coolant systems or containment which are not required to be open during accident conditions are closed. These valves may be opened to perform i

necessary operational activities.

2.

At least one door in each airlock is closed and sealed.

3 All automatic containment isolation valves are operable or deactivated in the isolated position.

4.

All blind flanges and manways are closed.

P.

Secondary Containment Integrity - Secondary containment integrity means that the reactor building is intact and the following conditions are met:

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b TABLE 3.1.A h

REACTOR PROTECTION SYSTEM (SCRAM) INSTR',PINTATIO:; REQUIREMEST Min. No.

Operable Inst.

Modes in Which Function h

Cha nnels Nst Be Operable Per Trip Shut-Startup/ Hot S yr.t o m (1)

Trio Function Trip Level Settint down Refuel (7)

Standby M

g,gg,ngg) f 1

Mode Ovitch in Shutdown g

1 Manual Scram I

X X

X 1.A IPM (16) 3

!!!gh Flux 1 IQ/l y ndicated

?-(22) I (22)

X (5) 1A 3

Inoperative I

X (5) 1A w

APP.M (16) 2 High Flux See Spec. 2.1.A.1 2

Hic;h Flux 115% rated power x

1.A or 1.3 2

Inoperative (13)

X(21)

X(17)

(15) 1.A or !.3 2

Downscale 1 3 Indicated on Scale X(21)

X(17)

X 1 ^ OF 1 3 l

(11)

(11)

X(12) 1.A or 1.8 2

Hig:s Reactor Pressure < 1055 psig X(10)

X X

l'A 2

Hic.h Drvvell

- 2.5 psig X(8)

X(6)

X l.A Pressure (14) 2 Reactor Lov Water

> 538 above vessel aero Level (14)

X X

X 1.A

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2 Hieh *.iater Level in Scram

< 50 Cellons X

Discharge Tank X(2) y K

1.A i

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NOTES FOR TABLE 3.2.P (1) From and af ter the date that one of these parameters is reduced to one indication, continued operation is permissible during the succeeding thirty days unless such instrumentation is sooner made operable.

(2) From and af ter the date that one of these parameters is not indicated in the control room, continued operation is permissible during the succeeding seveo days unless such instrumentation is sooner made operable.

(3) If the requirements of noces (1) and (2) cannot be met, and if one of the indications cannot be restored in (6) hours, an orderly shutdown shall be initiated and the reactor shall be in a cold condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

(84) These surveillance instruments are considered to be redundant to each other.

(5) If the requirements of notes (1) and (2) cannot be met, and if one of the indications cannot be restored in six (6) hours, an orderly shutdown shall be initiated and the reactor shall be in a Cold Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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3 3/4.3 BASES :

The surveillance requirement for scram testing of all the control rods after each refueling outage and 10% of the control rods at 16-week intervals is adequate for determining the operability of the control rod system yet is 1

not so frequent as to cause excessive wear on the control rod system components.

The numerical values assigned to the predicted scram performance are based on the analysis of data from other BWR's with control rod drives the same as those on Browns Ferry Nuclear Plant.

The occurence of scram times within the limits, but significantly longer than the average, should be viewed as an indication of systematic problem with control rod drives eSpecially if the number of drives exhibiting such scram times exceeds eight, the allowable number of inoperable rods.

In the analytical treatment of the transients which are assumed to scram on j

high neutron flux, 290 milliseconds are allowed between a neutron sensor reaching the scram point and the start of control rod motion.

This is adequate and conservative when compared to the typical time delay of about 210 milliseconds estimated from scram test results. Approximately tie first 90 milliseconds of each of these time intervals result from sensor atA circuit delays after which the pilot scram solenoid deenergizes to 120 milliseconds later, the control rod motion is estimated to actually begin.

However, 200 milliseconds, rather than 120 milliseconds, are conservatively assumed for this time interval in the transient analyses and are also included in the allowable scram insertion times of Specification 3.3.c.

• In order to perform scram testing as required by specification 4.3.c.1, the relaxation of certain restraints in the rod sequence control system is required. Individual rod bypass switches may be used as described in specification 4.3.c.1.

The position of any rod bypassed must be known to be in accordance with rod withdrawal sequence. Bypassing of rods in the manner described in specification 4.3.c.1 will allow the subsequent withdaawal of any rod scrammed in the 100 percent to 50 percent rod densitry groups; however, it will maintain group notch control over all rods in the 50 percent density to preset power level range. In addition, RSCS will prevent movement of rods in the 50 percent density to preset power level range until the secammed rod has been withdrawn.

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9 3.11 BASES The High Pressure Fire and CO, Fire Protection specifications are provided in order to meet the preestablished levels of operability during a fire in either or all of the three units. Requiring a patrolling fire watch with portable fire equipment if the automatic initiation is lost will provide (as does the automatic system) for early reporting and immediate fire fighting capability in the event of a fire occurrence.

The High Pressure Fire Protection System is supplied by four pumps (three electric driven and one diesel driven) aligned to the high pressure fire header. The reactors may remain in operation for a period not to exceed 7 days if three pumps are out of service. If at least two pumps are not made operable in seven days or if all pumps are lost during this seven day period, the reactors will be placed in the cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

For the areas of applicability, the fire protection water distribution system minimum capacity of 2664 gpm at 250' head at she fire pump discharge consists of the following design loads:

2 2

1.

Sprinkler System (0.30 gpm/ft /4440 ft area) 1332 gpm 2.

1 1/2" Hand Hose Lines 200 gpm 3

Raw Service Water Load 1132 ap; TOTAL 2664 gpm The CO, Fire Protection System is considered operable with a minimum of 8 1/2 tons (0.5 tank) CO in storage for units 1 and 2; and a' minimum of 3 tons 2

(0.5 tank) CO in storage for unit 3.

An immediate and continuous fire watch 2

in the cable spreading room or any diesel generator building area will be established if CO fire protection is lost in this room and will continue 2

until CO fire protection is restored.

2 To assure close supervision of fire protection system activities, the remo~al from service of any component in either the High Pressure Fire System or the C0 Fire Protection System for any reason other than testing or emergency p

operations will require Plant Superintendent approval.

Early reporting and immediate fire fighting capability in the event of a fire occurrence will be provided (as with the automatic system) by requiring a patrolling fire watch if more than one detector for a given protected zone is inoperable.

A roving fire watch for areas in wh.ch automatic fire suppression systems are to be installed will provide additional interim fire protection for areas that have been determined to need additional protection.

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UNIT 3

. PROPOSED CHANGES 1

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Ho t_ _ St an dby Condition - Hot standby condition means operation with coolant temperature greater than 2120F, system pressure less than 1055 psig, the main steam isolation valves closed and the mode switch in the Startup/ Hot Standby position.

J.

Cold Condition - Reactor coolant temperature equal to or less than 2120F.

K.

Hot Shutdown - The reactor is in the shutdown mode and the reactor coolant temperature greater than 2120F.

L.

Cold shutdown - The reactor is in the shutdown mode, the reactor coolant temperature equal to or less than 2120F, and the reactor vessel is vented to atmosphere.

M.

Mode of Operation - A reactor mode switch selects the proper interlocks for the operational status of the unit. The following are the modes and interlocks provided:

1.

Startup/ Hot Standby Mode - In this mode the reactor protection scram trips initiated by condenser low vacuum and main steam line isolation valve closure, are bypassed when reactor pressure is less than 1055 psig, the reactor protection system is energized with IRM neutron monitoring system trip, the APRM 15% high flux trip, and control rod withdrawal interlocks in service.

This is often ref erred to as just Startup Mode.

This is intended to imply the Startup/ Hot Standby position of the mode switch.

2.

Run Mode - In this mode the reactor system pressure l

is at or above 825 psig and the reactor protection system is energized with APRM protection (excluding i

the 15% high flux trip) an3 RBM interlocks in service.

l 3.

Shutdown Mode - Placing the mode switch to the shutdown position initiates a reactor scram and power to the control rod drives is removed.

After i

a short time period (about 10 sec), the scram i

signal is removed allowing a scram reset and restoring the normal valve lineup in the control I

rod drive hydraulic system; also, the main steam line isolation scram and main condenser low vacuum scram are bypassed if reactor vessel pressure is below 1055 psig.

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TABLE 3.1.A REACTOR PROTDCTION SYSTD4 (SCRAM) INSTRIB(ENTATION EEQUIRDtDt?

Min. No.

of operable Modes in Which Function Inst.

Munt fu operable mannels Shut-S ta rt up/ Dot

,Per Trip

,Systes fil Trio Function Trip Ipvel Setting down Refuel 17)

Standby Etg}

Act ion f 11 1

Mode Switch in Shutdown I

1 X

X 1.A X

X X

X 1.A 1

Manual Scram IRM (16) 3 Bigh Flux 5 120/125 Indicated on scale x(22) X (22) x (5) 1.A X

X (5) 1.A 3

Inoperative APRM (16) x 1.A r 1.B 2

Bigh Flux See Spec. 2.1.A.1 (21) x(17)

(15) 1.A or 1.B 2

Bigh Flux 5 151 rated power X

ti 2 Inoperative (13)

X (21)

Z(17)

I 1.A or 1.B 2

Downscale 2 3 Indicated on Scale (11)

(11)

I(12),

1.A or 1.8 2

High Reactor Pressure 5 1055 psig Z(10)

X X

1.A 2

High Drywell Pressure (14) s 2.5 psig x (8)

Z(8)

X 1.A 2

Reactor Low Water Level (14) 2 518= above vessel zero X

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1.A 2

High Water I* vel in Scram X*

1.A Discharge Tank 5 50 Gallons 1

X(2)

X 4

Main Steam Line Isola-tion valve closure 5105 valve closure x (3) (6) x (3) (6). x(6) 1.A or 1.C 2

Turbine cont. valve Upon trip of the fast Fast Closure acting solenoid valves E (1)

I(4)

Z(4) 1.A or 1.D e.

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NOTES FOR TABLE 3.2.F (1) From and af ter the date that one of these parameters is reduced to one indication, continued operation is permissible during the succeeding thirty days unless such instrumentation is sooner made operable.

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(2) From and af ter the date that one of these parameters is not indicated in the control room, continued operation is permissible during the succeeding seven days unless such 1

instrumentation is sooner made operable.

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(3) If the requirements of notes (1) and (2) cannot be met, and if one j

of the indications cannot be restored in (6) hours, an orderly shutdown shall be initiated and the reactor shall be in a cold condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

j (4) These surveillance instruments are considered to be redundant i

to each other.

(5) If the requirements of notes (1) and (2) cannot be met, and if one of the indications cannot be restored in six (6) hours, an orderly shutdown shall be initiated and the reactor shall be in a Cold Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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In the analytical treatment of the transients which are assumed to scram on high neutron flux, 290 milliseconds are allowed between a neutron sensor reaching the scram point and the n te. r t of control rod motion.

Thin is adequate and conservative when compared to the typical ti.mc delay of about 210 milliseconds estimated f rom scram test result.).

Approximately the first 90 milliseconds of each of these time intervals result from the sensor and circuit delays af ter which the pilot scram solenoid deenergi::es and 120 n1111 seconds later, the control rod ration is estimated to actually begin, llowever, 200 milliseconds,rather than 120 milliseconds,are conservatively assumed for this time interval in the transient analyses and are also included in the allowable scram insertion times of Specification 3.3.C.

In order to perform scram time testing as required by specifiestien

1. 3.C.1, the relaxation of certain restraints in the rvd cequence control sy tem is required. Individual rod bypass switches may be used as described in specification' h.3.C.l.

The position of any rod bypassed must be known to be in accordance with rod withdrawal sequence. Bypassin6 of redc in the conner deceribed in specification L.3.C.1 will allow the subsequent withdrawal of any rod scrac:ned in the 100 percent to 50 percent rod density groups; however, it will maintain group notch control over all rods in the 50 percent to o percent rod density aroups. In addition. RSCS will prevent movemenr.

of rods in the 50 percent density to a preset power level range until the scrammed rod has been withdrawn.

D.

Reactivity Anomalies Durino each fuel cycle excess operative reactivity varies as iuel depletes and as any burnable poison in supplementary control is burned.

The magnitude of this excess reactivity may be inferred from the critical rod configuration.

As fuel burnup progresses, anomalous behavior in the excess reactivit y may be detected by comparison of the critical rod pattern at selected base states to the predicted rod inventory at that state.

Power operating base conditions provide the most sensitive and directly interpretable data relative to core reactivity.

Furthermore, using power ope ra ting base conditions permits f requent reactivity compa ri sons.

Requiring a reactivity comparison at the specified frequency annures that a comparison will be made before the core reactivity change exceeds 1% d K.

Deviations in core reactivity greater than 17. d K are not expected and require thorough evaluation.

One percent reactivity limit is concidered safe since an insertion of the reactivity into the I

core would not lead to transients exceeding design conditions o t' the reactor system.

References Gener'l Electric Supplemental Reload Licensin5 Submittal for j

1.

a EFI;P unit 3 Reload 2, NEDo-24199, July 1979.

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3.11 BASES The High Pressure Fire and C0 Fire Protection specifications are provided 7

in order to meet the preestablished levcls of operability during a fire in either or all of the three units. Requiring a patrolling fire watch with portable fire equipment if the automatic initiation is lost will provide (as does the automatic system) for early reporting and immediate tire fighting capability in the event of a fire occurrence.

The High Pressure Fire Protection System is supplied by four pumps (three electric driven and one diesel driven) aligned to the high pressure fire header. The reactors may remain in operation for a period not to exceed 7 days if three pumps are out of service. If at least two pumps are not made operable in seven days or if all pumps are lost during this seven day period, the reactors w'.ll be placed in the cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

For the areas of applicability, the fire protection water distribution system minimum capacity of 2664 gpm at 250' head at the fire pump discharge consists of the following design loads:

2 2

1.

Sprinkler System (0 30 gpm/ft /4440 ft area) 1332 gpm 2.

1 1/2" Hand Hose Lines 200 gpm 3

Raw Service Water Load 1132 gpm TOTAL 2664 gpm The CO, Fire Protection System is considered operable with a minimum of 8 1/2 tons (0.5 tank) C0 in storage for units 1 and 2; and a minimum of 3 tons p

(0.5 tank) CO in storage for unit 3.

An immediate and continuous fire watch 2

in the cable spreading room or any diesel generator building area will be established if CO fire protection is lost in this room and will continue 2

until CO fire pr tection is restored.

2 l

To assure close supervision of fire protection system activities, the removal Trom service of any component in either the High Pressure Fire System or the C0 Fire Protection System for any reason other than testing or emergency 7

operations will require Plant Superintendent approval.

Early reporting and immediate fire fighting capability in the event of a fire occurrence will be provided (as with the automatic system) by requiring a patrolling fire watch if more than one detector for a given protected zone is inoperable.

. A roving fire watch for areas in which automatic fire suppression systems are to be installed will provide additional interim fire protection for areas that have been determined to need additional protection, 356 i

ENCLOSURE 2 JUSTIFICATION Pages Reason for Change Appendix A Units 1 and 2, page 4 Brings mode switch run mode definition unit 3, page 3 value of minimum system pressure into alignment with paragraph 2.1.0 and H, Bases.

Units 1 and 2, page 33 The High Drywell Pressure switches are unit 3, page 32 shared between RPS (table 3 1.A) and Primary Containment (table 3.2.A).

In table 3.2. A the setpoint for this switch is f 2 5 psig, therefore, the setpoint in RPS should also be f 2.5 psig.

Units 1 and 2, page 80 Deletes erroneous reference in existing note 3 unit 3, page 83 to technical specification section 3 5.H.

Units 1 and 2, page 133 The transient analyses for all three Browns unit 3, page 136 Ferry units have used a value of 290 milli-seconds for transients which are assumed to scram on high neutron flux. This 290 milliseconds is the time from the neutron sensor reaching set-point to the start of control rod motion.

Units 1 and 2, page 326 Reflects the addition of a diesel driven fire unit 3, page 356 pump.

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