ML20086E232

Proposed Tech Specs Section 3.1/4.1, Reactor Protection Sys

ML20086E232
Person / Time
Site:	Quad Cities
Issue date:	11/20/1991
From:	COMMONWEALTH EDISON CO.
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ML20086E228	List: ML20086E227 ML20086E232
References
NUDOCS 9111270109
Download: ML20086E232 (45)

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O PROPOSED TECH SPEC TS 3.1/4.1

, ' REACTOR PROTECTION SYSTEM' e .

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QUAD CITIES UNITS 1 & 2 I DPR-29 & DPR-30 l 0 3.1/4.1 REACTOR PROTECTION SYSTEM SPECIFICATIONS ,

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS O A. Reactor. Protection System A. Reactor Protection System The reactor protection system Surveillance of the reactor instrumentation CHANNELS shall be protection system instrumentation '

OPERABLE with the setpoints, CHANNELS shall be performed as; minimum number of - TRIP SYSTEMS follows: ,

O and minimum number of instrument t C}lANNELS as shown in Table 3.1-1. 1. Reactor protection The system response times from instrumentation systems the opening of the sensor contact shall be functionally 1 up tocand including the opening tested, calibrated and, of the trip actuator contacts checked as indicated in'

~O shall not exceed so milliseconds. Tables 4.1-1 and 4.1-2.  ;

APPLICABILITY: 2. The system response times for each Trip Function shown t 3 As shown in Table 3.1-1. in Table 3.1-1 shall bel demonstrated to be within'

~O ACTIONt its limit at least .eachl REFUELING OUTAGE. Each test ;

1. . . With a reactor protection shall include at least one' system instrumentation - CilANNEL per TRIP SYSTEM such .

setpoint less conservative that all CHANNELS are tested i than the value shown in the- at least once every (N)

O' Trip Level Setting column of - REFUELING OUTAGES where (N)

Table 3.1-1, declare the is the total number ofi CHANNEL !noperable and redundant CHANNELS in ai follow ACTION 3.1.A.2 or specific reactor TRIPj 3.1.A.3 below until the SYSTEM. ,

CHANNEL is restored to O OPERABLE status with its setpoint adjusted. consistent-  ?

with the Trip Level Setting value.  ;

2.- With the number of OPERABLE *

'O' CHANNELS less-than required-by the -Minimum OPERABLE .

CHANNELS per TRIP SYSTEM requirement for one TRIP ,

SYSTEM, place the inoperable CHANNEL (s) and/or that TRIP SYSTEM- in the tripped O

3.1/4.1-1

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QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 g

condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

An inoperable CHANNEL need not be placed in the tripped condition when this would cause the PROTECTIVE g

FUNCTION to occur. In these cases, the inoperable CHANNEL shall be restored to OPERABLE status within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after the CHANNEL was first determined to be inoperable or the ACTION required by Table 3.1-1 shall be entered.

3. With the number of OPERABLE CHANNELS less than required g

by the Minimum OPERABLE CHANNELS por TRIP SYSTEM requirement for both TRIP SYSTEMS, place at least one TRIP SYSTEM in the tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and I then take the ACTION required by Table 3.1-1.

The TRIP SYSTEM need not be placed in the tripped condition. if this would cause the PROTECTIVE I WNCTION to occur. When a TRIP SYSTEM can be placed in the tripped condition without causing the PROTECTIVE FUNCTION to occur, place the TRIP SYSTEM I with the most inoperable CHANNELS in the tripped condition; if both systems have the same number of inoperable CHANNELS , place either TRIP SYSTEM in the

> tripped condition.

B. APRM Scram and Control Rod Block B. APRM Scram and Control Rod Bloc Flow Biased UpsF:.a l Setpoints Flow Biased Upscale Setpoints The APRM flow bit sed neutron flux The core power distribution shal upscale scram trip setpoint and be checked daily for MFLPD ar flow biased neutron flux upscale compared with the FRP.

3.1/4.1-2

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QUAD CITIES UNITS 1 & 2

< DPR-29 & DPR-30 7 O  ;

i- control rod block trip setpoint

! shall be established according to j- the equations in Specifications ,

2.1.A.1 and-2.1.B.

[O APPLICABILITYt 1 i j OPERATIONAL MODE 1, when thermal

power is greater than or equal to

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) 25% of RATED THERMAL POWER.

20- -ACTION

! l'. With- the APRM flow biased i neutron flux upscale scram trip setpoint- and/or the flow- biased neutron flux

0 upscale control rod block

]

trip .setpoint- less

conservative than the value shown in the equations in i i,.

Specifications 2.1.A.1 and 2.1.B, initiate corrective lO

action within 15 minutes and -

i within 6. hours, adjust the

setpoints to be consistent '

i with the Trip Setpoint

] -values.or increase the APRM i gain as described in f0 Specification 2.1.A.1 and '

E 2.1.B or reduce thermal.

l power'to less than 25% of i RATED THERMAL POWER within

the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

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QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 j TABLE 3.1-1 REACTOR EEQ2ICTION SYSTEM fSCRAM) INSTRUMENTATION R EOUIREMENTS Minimum OPERASLE Applicable ,

CHANNELS Per Trip Level OPERATIONAL Trip runction TRIP SYSTEM (a)(b) Setting MODES ACTION

~~~~-~~~-------~~~~~~~-----~~'---~~~~~~~~~~~~~-~~'~~~~~~-~~----~~--~~-----~~~~~-~~~~~

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1. Mode Switch 1 N.A. 1, 2 1 L in Shutdown 1 3, 4 7 1 5 2

2. Manual Scram 1 N.A. 1, 2 1 ,

O 1 3, 4 7 1 5 6

3. IRM (c)  ;

a. High Flux 3 $ 120/125 of full 2 1  ;

2 scale 3, 4 7 0 3(n) 5(m) 2

b. Inoperative 3 N.A. 2 1 2 3, 4 7 3(n) 5(m) 2 O 4. APRM (f)

a. High Flux 2 Tech Spec 2.1.A.1 1 3 (flow biased)

b. Inoperative 2 N.A. 1, 2 1_  ;

2 3 7 0 2(n) 5(m) 2

c. High Flux- 2 Tech-Spec 2.1.A.2 2 1 3 7 (154 scram) 2 5(m) 2 2(n) ,

d. High Flux 2 Tech Spec 2.1.A.1 1 3 "O- (Scram Clamp)

5. Reactor High 2 5 1060 psig 1, 2(g) 1-Pressure.

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6. Drywell_High 2 s 2.5:poig 1, 2(h) 1.

O. Pr..sur.

7. Reactor Low 2 >

,,8 inches'(d) 1, 2 1 ,

Water Level 3.1/4.1-4 0

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• QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 TABLE 3.1 _ REACTOB PROTECTION SYSTEM fSCRAM) INSTRUMENTATION REOUIREMENTS Minimum OPERASLE -

Applicable CHANNELS Per Trip Level OPERATIONAL Trip Function TRIP SYSTEM (a)(b) Setting MODES ACTION D ..................................................................... ____........

8. Scram Discharge 2/ bank 5 40 gallons 1, 2 1 Volume High 2/ bank 5(i)(1) 2 Water Level

} 9. Turbine condenser 2 1 21 inches Hg 1 4 Low Vacuum vacuum

10. Main Steam Line 2 1 15 X Normal Full 1, 2(g) 4 l'

?. High Radiation- Power Sackground (e) 3 18. . Main Steam Line 4 (k) $ 10% Valve 1 3 Isolation valve closure closure

18. Turbine control 2 3 460 poig (o) 1(j) 5 Valve Fast Closure-

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13. Turbine Stop 4 5 10% Valve 1(j) 5 Valve closure closure

- 14 . _ Turbine EHc 2 2 900 peig 1(j) 5 control Fluid O Low Pressure t

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QUAD CITIES UNITS 1 &2 DPR-29 & DPR-30 0 TABLE 3.1-1 (continued)

REACTOR PROTECTION SYSTEM fSCRAM) INSTRUMENTATION REQUIREMENTS ACTIONS

() ACTION 1 - Be in at 1 cast HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

ACTION 2 - Suspend all operations involving CORE ALTERATIONS

• and insert all insertable control rods within one hour.

() ACTION 3 - De in at least STARTUP within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

ACTION 4 - De in STARTUP with the main steam line isolation valves closed within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

O ACTION $ - Initiate a reduction in thermal pcwor within 15 minutes and reduce turbine first stage pressure to that which corresponds to less than 45% of rated steam flow, within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

ACTION 6 - Suspend all operations involving CORE ALTERATIONS *,

() and insert all insertable control rods and lock the reactor mode switch in th9 SHUTDOWN position within one hour.

ACTION 7 - Verify all insertable control rods to be inserted in the core and lock the reactor mode switch in the

() Shutdown position within one hour.

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• Except replacement of LPRM strings provided SRM CF instrumentation is OPERABLE per Specification 3.10.B.

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QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 O

TABLE 3.1-1 (Continued)  ;

t REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REOUIngll1{If5 TABLE NOTATIONS g

(a) CHANNEL may be placed in an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for I required surveillance without placing the TRIP SYSTEM in the tripped condition provided at least one OPERABLE CHANNEL in the same TRIE!

SYSTEM is monitoring that parameter.

O- (b) Two TRIP SYSTEMS shall be OPERABLE in the applicable OPERATIONAI MODES for the specified Trip Function. C11ANNEL OPERABILITY requirements within the T:< I P SYSTEM are specified in the ACTION provisions of Specification 3.1.A.

(c) This function shall be automatically bypassed when the reactor modc g switch is in the RUN position.

(d) The +8-inch trip point is the water level as measured by the-instrumentation outside the shroud. The water level inside the shroud will decrease as power is increased to 100% in comparison tc the level outside the shroud to a maximum of 7 inches. This is duc O to the pressure drop across the steam dryer. Therefore, at 1004 power, an indication of +8 inch water level will actually be +1 inc!.:

inside the shroud. 1 inch on the water level instrumentation is 2 504" above vessel zero.

(e) CHANNEL shared by the reactor protection and containment isolatior '

d- system.

(f) An APhtM will be considered inoperable if there are fewer than 2 LPPJ '

l inputs per level or there are less than 50% of the normal complement of LPRMs to an APRM.

O (g) This function is not required to be OPERABLE when the reactol' pressure vessel head is not bolted to the vessel.

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

f k (i)' With'any control rod withdrawn. Not applicable to control rods removed per Specification 3.10.D or 3.10.E.

(j) Permissible to bypass when turbine first stage pressure is less than that which corresponds,to 45% of rated steam flow (< 400 psi).

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QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 0

TABLE 3.1-1 (Continued)

PEACTOR PROTECTION SYSTEM fSCRAM) INSTRUMENTATION REOUIREMENTS TABLE NOTATIOt!S*

C)

(k) The design permits closure of any one line without a scram beins' initiated.

(1) Permissible to bypass, with control rod block, for reacto:

protection system reset in REFUEL and SHUTDOWN ?ositions of th(t

() reactor mode switch.

(m) The " shorting links" shall be removed from the RPS circuitry prior :

to and during the time any control rod is withdrawn and shutdowr>

margin demonstrations are being performed. Not required for contro):

rods removed per Specification 3.10.D or 3.10.E.

(n) The non-coincident NMS reactor trip function logic is such that al]l channels go to both trip systems. Therefore, when the "shortins, links" are removed, the Minimum OPERABLE CHANNELS Por TRIP SYSTEM 11:

4 APRMS and 6 IRMS.

C) (o) Trip is indicative of turbino control valve f ast closure (due to lot ;

EHC fluid pressure) as a result of fast acting valvo actuation.

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QUAD CITIES U!1ITS 1 & 2 DPR-29 & DPR-30 TABLE 4.1-1 g

REACTOR PPQTECTION SYS_ TEM fSCPAM) INSTRUMENTATION CHANNEL FUNCTIONAL TEST AND CHANNEL CJ1ECK PEOUIREMENTS g ...................................................................................

CHANNEL CHANNEL Applicable TUNCTIONAL TUNCTIONAL CHANNEL OPERATIONAL ,

TEST Method TEST CHECKS MODES Trip Tunction (a) (c)

NA 1, 2, 3, 4, 5

) 1. Mods Switch in Place Mode Sw R Snutdown in Snutdown W NA 1, 2, 3, 4, 5

2. Manual Scram Trip Channel and Alarm

3. IRM g

a. High Flux Trip Channel S/U(e), W S/U, S, (b) 2(k) and Alarm (d) W S 3, 4, 5 2(k), 3, 4, 5

b. Inoperative Trip Channel W NA and Alarm D 4. APRM Trip Output W(i), Q S, D(1) 1

a. High Flux (flow biased) Relays (d)

NA 1, 2 , 3, 5

b. Inoperat ive Trip Output Q Relays

)

c. High Flux Trip Output S/U(e), W S/U, S, (b) 2(k)

Relays (d) W S 3, 5 (15% scram)

W(i), Q S 1

d. High Flux Trip Output (Scram Clamp) Relays (d) g 1, 2(h)

5. Reactor High Trip Channel Q NA Pressure and Alarm Trip Channel NA 1, 2(j)

6. Drywell High Q Pressure and Alarm 9 7. Reactor Low Trip Channel Q D 1, 2 Water Level and Alarm 9

3.1/4.1-9 b .

O. 1 QUAD CITIES UllITS 1 & 2 DPR-29 & DPR-30 O' TABLE 4.141 REACTOR PROTECTION SYSTEM fSCRAM) If1STRUMENTATION CHANilEL TUNCTIOllAL TEST AND CHANNEL CHECK REOUI.REMENTS O j CHANNEL CHANNEL Applicable i TUNCTIONAL TUNCTIONAL CRANNEL CPERATIONAL '

Trip runction T1,87 Methods TEST CHECK MODES (a) (c) 8.= scram Discharge Trip Channel Q NA 1, 2, 5(g) volume-High Water .and Alarm (f) ,

Level (Thermal and dp switch)

9. Turbine Condenser Trip Channel Q NA 1  ;

O -Low vacuum- and Alarm ,..

t Trip Channel Q 5 1, 2(h)

10. . Hain steam Line ,

High Radiation. and Alarm (d)

11. Main steam Line Trip Channel Q NA 1 O- Isolation Valv. and Alarm Closure i

'12. Turbing Control Trip Channel Q NA 1 Valve Fast Closure and Alarm l

13. Turbine stop Trip Channel Q NA 1 10 Valve Closure and Alarm

14. -Turbine EHC Control Trip Channel Q NA - 1 Fluid Low Pressure and Alarm ,

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O I QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30  ;

TABLE 4.1-1 (Continued)

REACTOR PROTECTION SYSTEM fSCRAM) INSTRUMENTATIOff ,

CHANNEL PUNCTIONAL TEST AND CHANNEL CHECK REOUIREMENTS ,

T6 ALE NOTATIONS l (a) A CHANNEL FUNCTIONAL TEST of the logic of each CHANNEL is performeti as indicated. This coupled with placing the mode switch in Shutdow:!

sach REFUELING OUTAGE constitutes a LOGIC SYSTEM FUNCTIONAL TEST o !

g the scram system.

(b) The IRM and SRM channels shall be determined to overlap for at leas-(1/2) decades during each startup af ter entering OPERATIONAL MODE .,

and the IRM and APRM channels shall be determined to overlap for a" least (1/2) decadas during each controlled shutdown, if no perf rmed within the previous 7 days.

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(c) CHAlWEL FUNCTIONAL TESTS are not required when the systems are no required to be OPERABLE or are tripped. If tests are missed, the!

shall be-performed prior to returning the systems to:an OPERABLl status.

O (d) Tnis-instrumentation is exempte'd from the CHANNEL FUNCTIONAL TES definition (Definition 1.6). This CHANNEL FUNCTIONAL TEST wil; a simulated electrical signal into thi consist of . injecting

- measurement CHANNELS.

'(e) Within 24' hours prior. to atartup, if not performed - within - th;H

_O- previous 7 days.-

(f) only the ol'actronics portion of the thermal switches-will be teste!

using an electronic calibrator during the three month test. A wate:

column or. equivalent will be used to test the dp switches.

~O' (g) - With any control rod withdrawn. Not applicable to control rodi removed per Specification 3.10.D or 3.10.E.

(h)- This function is not required to be OPERABLE when the reacte pressure vessel head is not bolted to the vessel.

O' (i) Within one week after entering OPERATIONAL MODE 1 and then quartert thereaftier.--

(j) This function is- not required to be OPERABLE when PRIMAF

' CONTAINMENT INTEGRITY is not required.

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QUAD CITIES Ul1ITS 1& 2 DPR-29 & DPR-30 O TABLE 4.1-1 (Continuedi*

PEACTQR PROTECTIOfi SYSTEli_(f.CIVdi) IfiSTRUEEllTAT101{

CJIA!1tiEL FUllCTIOffAL TEST AllD CllA111(EL CJ1ECK RPOUIREME11TE O TAatE riOTATIOriS (k) The provisions of Specification 4.0.D are not applicable providet the CilAllliEL FUliCTIOffAL TEST is performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> aftet entering OPERATIOl{AL MODE 2 from OPERATIO!iAL 140DE 1.

C (1) Verify measured core flow to be greater than or equal to establishec core flow at the existing pump speed.

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• QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 TABLE 4.1-2 ,

i EACTQE PROTECTIOf{ SL911M f SCRAM) ,U(ETitilMENTATION QlAlulU d LIBRATION DIQUIREMENTS CHANNEL Applicable CALIBRATION Minimum OPERATIONAL

,0 Trip runction Method (a) (f) Frequency (b) MODES

1. IRM High Flux Electronic R 2 Calibration R 3, 4, 5

2. APRM High Flux >

a. Flow slas standard Pressure W(d)(k), R 1 .

and Voltage source  !

b. 15% Scram Electronic R 2 Calibration R 3, 5 O'

c. Scram Clamp Electronic W(d), R 1 Calibration ,

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3. LPRM (h) Using TIP System (g) 1

4. Reactor High Standard Pressure Q, 1, 2(j)

O pressure source G. Drywell High Standard Pressure Q 1, 2(1) '

Pressure source

6. Reactor Low standard Pressure R(e) 1, 2

,Q Water Level source

7. Turbina condenser standard Vacuum source Q 1 Low Vacuum ,

8. Main steam Line Appropriate Radiation- R 1, 2(j)

High Radiation. Source (c)

9. Turbine EHC Control Standard Pressure Q 1 Fluid Low Pressure Source

10. Turbine Control Valve Standard Pressure R 1 Fast Closure Source

'O 11. High Water Level in Standard Pressure R 1, 2, 5 scram Discharge source Volume (dp only) 3.1/4.1-13 0

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l QUAD CITIES UNITS 1 & 2 U""~*' ' ""~'

.O l TABLE 4.1-2 (Continued)

R..EACTOR PROTECTION SYSTEM fSCRAM) INSTRUMENTATION REOUIREMENTS

- CHANNEL CALIBRATION REOUIREMENTS TABLE NOTATIONS I 1

(a) Neutron detectors may be excluded from the CHANNEL CALIBRATION.

(b) CHANNEL CALIBRATION tests are not required when the systems are not:

O required to be OPERABLE or are tript ed. If tests are missed, theyi shall be performed prior to rc* urning the systems to an OPERADLE status.

(c) A current source provides an instrument CHANNEL alignment overy 3 months.

(d) This calibration shall consist of the adjustment of the APRM channel; to conform to the power values calculated by a heat balance during.

OPERATIONAL MODE 1 when thermal power > 25% of RATED THERMAL POWER.

Adjust the APRM channel if the absolute difference is greater than!

2% of RATED THERMAL POWER. Any APRM channel gain adjustment made in O compliance with Specification 2.1.A or 2.1.B shall not be included' in determining the absolute difference.

(e) Trip units are calibrated at least once per quarter and transmitters are calibrated at least once per_OPEPATING CYCLE.

O (f). Response time is not part of the routine CHANNEL CHECK and CHANNE1 CALIBRATION.

(g) Every 1000 equivalent full power hours.

(h) Does not provide scram function.

(i) This -function is not required to be OPERABLE when PRIMAR(

CONTAINMENT INTEGRITY is not required.

(j) This function is not required to be OPERABLE when the reactor!

pressure vessel head is not bolted to the vessel.

(k) This-Calibration shall consist of the adjustment of the APRM floi biased channel to conform to a calibrated flow signal.

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QUAD CITIES UNITS 1 &2 DPR-29 & DPR-30 0

3.1 LIMITING CONDITIONS FOR OPERATION BASES The reactor protection system automatically initiates a reactor scram to:

O a. preserve the integrity of the fuel cladding,

b. preserve the integrity of the primary system, and

c. minimize the energy which must be absorbed and prevent criticality following a loss-of-coolant accident. ,

O This specification provides the LIMITING CONDITIONS FOR i OPERATION necessary to preserve the ability of the system to tolerate single failures and still perform its intended function, even during periods when instrument channels may be out-of-service because of maintenance. When necessary, one g channel may be made inoperable for brief intervals to conduct required fianctional tests and calibrations.

The reactor protection system is of the dual channel type

. (reference SAR Section 7.7.1.2) . The system is made up of two independent trip systems, each having two subchannels of g tripping devices. Each subchannel has an input from at least' one instrument channel which monitors a critical parameter.

The outputs of the subchannels are combined in a one-out-of-two-logic, i.e., an input signal on either one or both of the subchannels will cause a trip system trip. The outputs of the trip systems are arranged so that a trip on both systems O is required to produce a reactor scram.

This system meets the requirements of the IEEE 279, " Standard for Nuclear Power Plant Protection Systems" issued September 13, 1966. The system has a reliability greater than that of a tw - ut- f-three system and somewhat less than that of a O one-out-of-two system (reference APED 5179).

With the exception of the average power range monitor (APRM) and intermediate range monitor (IRM) channels, each subchannel has at least one instrument channel. When the minimum c nditi n f r perati n n the number of operable instrument O channels per untripped protection trip system is met, or if it cannot be met and the affected protection trip system is placed in a tripped condition, the effectiveness cf the protection system is preserved, i.e. , the system can tolerate a single failure and still perform its intended function of scramming the reactor. Three APRM instrument channels are O- provided for each protection trip system.

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QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 0

APRMs #1 and $3 operate contacts in one subchannel and APRMs 42 and #3 operate contacts in the other subchannel. APRMs 6 4, f 5 and #6 are arranged similarly in the other protection trip systum. Each protection trip system has one more APRM tnan is necessary to meet the minimum number required per channel .

O This allows the bypassing of one APRM por protection trip system for maintenance, testing, or calibration. Additional IRM channels have also been provided to allow for bypassing of one such channel. The bases for the scram settings for the IRM, APRM, high reactor pressure, reactor low water level, turbine control valve fast closure, and turbine stop valve O closure are discussed in Specifications 2.1 and 2.2.

Pressure ser. sing of the drywell is provided to detect a loss-of-coolant accident and initiate the emergency core cooling equipment. The pressure-sensing instrurentation is a backup to the water-level instrumentation which is discussed O in Specification 2.1. A scram is provided at the same setting as the amargency core cooling system (ECCS) initiation to minimize the energy which must be accommodated during a loss-of-coolant accident and to prevent the reactor from going critical following the accider.t.

O The control rod drive scram system in designed so that all of the water which is discharged from the reactor by a scram can be accommodated in the discharge piping. A part of this system is an individual instrument volume for each of the south and north CRD accumulators. These two volumes and their piping can hold in excess of 90 gallons of water and are the O low point in the piping. No credit was taken for these volumes in the design of the discharge piping relative to the amount of water which must be accommodated during a scram.

During normal operations, the discharge volumes are empty; .

however, should either volume fill with water, the water discharged to the piping from the reactor may not be accommodated which could result in slow scram times or partial O or no control rod insertion. To preclude this occurrence, level switches have been installed in both volumes which will alarm and scram the reactor wt.en the volume remaining in either instrument volume is approximately 40 gallons. For diversity of level sensing methods that will ensure and o provide a scram, both differential pressure switches and thermal switches have been incorporated into the design and logic of the system. The setpoint for the scram signal has been chosen on the basis of providing sufficient volume remaining to accommodate a scram, even with 5 gpm leakage per i drive into the SDV. As indicated above, there is sufficient l volume in the piping to accommodate the scram without 1 O impairment of the scram times or the amount of insertion of B 3.1/4.1-2 O l l

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QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 0

the control rods. This function shuts the reactor down while suf ficient volume remains to accommodate the discharged water and precludes the situation in which a scram would be required but not be able to perform its function properly.

O Loss of condenser vacuum occurs when the condenser can no longer handle heat input. Loss of condenser vacuum initiates a closure of the turbine stop valves and turbine bypass valves, which eliminates the heat input to the condenser.

Olosure of the turbine stop and bypass valves causes a pressure transient, neutron flux rise, and an increase in O surface heat flux. To prevent the cladding safety limit from being exceeded if this occurs, a reactor scram occurs on turbine stop valve closure. The turbine stop valve closure scram function alone is adequate to prevent the cladding safety limit from being exceeded in the event of a turbine trip transient with bypass closure.

O The condenser low-vacuum scram is a backup to the stop valve closure scram and causes a scram before the stop valves are closed, thus the resulting transient is less severe. Scram occurs at 21-inches Hg vacuum, stop valve closure occurs at 20-inches Hg vacuum, and bypass closure at 7-inches Hg vacuum.

O High radiation levels in the main staamline tunnel above that due to the normal nitrogen and oxygen radioactivity are an indication of leaking fuel. A scram is initiated whenever such radiation level exceeds fifteen times normal background (without hydrogen addition). The purpose of this scram is to reduce the source of such radiation to the extent necessary to O Discharge of prevent excessive turbine contamination.

excessive amounts of radioactivity to the site environs is prevented by the air ejector off-gas monitors, which cause an isolation of the main condenser of f-gas line provided the limit specified in specification 3.8 is exceeded. .

O The main steamline isolation valve closure scram is set to scram when the isolation valves are 10% closed from full open.

This scram anticipates the pressure and flux transient which would occur when the valves close. By scramming at this setting, the resultant transient is insignificant.

O A reactor mode switch is provided which actuates or bypasses the various scram functions appropriate to the particular plant operating status (reference SAR Section 7.7.1.2).

Whenever the reactor mode switch is in the REFUEL or STARTUP HOT STANDBY position, the turbine condenser low-vacuum scram O and main steamline isolation valve closure scrams are bypassed. This bypass has been provided for flexibility B 3.1/4.1-3 0 -

3, QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 e

during startup and to allow repairs to be made to the turbine condenser. While this bypass is in effect, protection is provided against pressure or flux increases by the high-pressure scram and APRM 15% scram, respectively, which are effective in STARTUP/ HOT STANDBY.

O If the reactor was taken to a hot standby condition for repairs to the turbine condenser, the main steamline isolation valves would be closed. No hypothesized single failure or single operator action in this mode of operation can result in an unreviewed radiological re?, case.

9 The manual scram function is active in all OPERATIONAL MODES, thus providing for a manual means of rapidly inserting control rods during all reactor OPERATIONAL MODES.

The IRM system provides protection against ex 've power 9 levels and short reactor periods in the rtup and intermediate power ranges (reference SAR Sections 7.4.4.2 and 7.4.4.3). During refueling, the primary Neutron Monitoring System (NMS) indication of neutron flux levels is provided by the Source Range Monitors (SRM). The SRMs provide input to the RPS, but shorting links are installed across the normally it closed contacts such that tripping an SRM channel does not affect the RPS. To activate the SRM scram function, these shorting links must be removed from the RPS. The SRM control rod scram provides backup protection to refueling interlocks and SHUTDOWN MARGIN should a prompt reactivity excursion occur. Although the IRM and APRM functions are required to be e OPERABLE during refueling, the SRMs provide the only on-scale monitoring of neutron flux levels during refueling and therefore the shorting links must be removed to enable the scram function of the SRMs. The RPS (and therefore removal of the RPS shorting links) is required to be OPERABLE in REFUEL only with any control rod withdrawn from a core cell containing one or more fuel assemblies. Control rods O withdrawn from a core cell containing no fuel assemblies do not affect the reactivity of the core and therefore are not required to have the capability to scram. Provided all control rods are otherwise inserted, the RPS function is not required. In this condition, the required SHUTDOWN MARGIN and

.g the one-rod-out interlock provide assurance that the reactor will not become critical thereby requiring a scram. If the SHUTDOWN MARGIN has been demonstrated, the RPS shorting links are not required to be removed. Under these conditivns, the capability of the one-rod-out interlock to prevent criticality has been demonstrated and the backup scram protection provided g by the IRMs is sufficient to ensure a highly reliable scram if required. In the power range, the APRM system provides B 3.1/4.1-4 D

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QUAD CITIES Uli1TS 1 & 2 DPR-29 & DPR-30 e

required protection (reference SAR Section 7.4.5.2). Thus, the IRM system is not required in the RUlf OPERATIOllAL MODE, the APRMs cover only the intermediate and power range r and the IRMs provide adequato coverage in the startup and intermediate range.

O The high reactor pressure, high drywell pressure, low reactor water level scrams are required for OPERATIOliAL MODES 1 and 2.

The scram discharge volume high level scram is required in OPERATIOliAL MODES 1, 2 and 5. They are therefore required to be operational for these OPERATIOllAL MODES of reactor e operation.

The turbine condenser low-vacuum scram is required only during power operation and must be bypassed to start up the unit.

4 O

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B 3.1/4.1-5 e

O, QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 0 4.1 SURVEILLANCE REQUIREMENTS BASES A. Surveillance requirements for the reactor protection system are selected in order to demonstrate proper function and operability. The surveillance intervals are determined in many different ways, such as, 1) operating O experience, 2) good engineering judgement, 3) reliability analyses, or 4) other analyses that are found acceptable to the NRC.

The f requency of calibration of the APRM flow-biasing netw rk has been established at each refueling outage.

O The flow-biasing network is functionally tested at least once per quarter and, in addition, cross calibration checks of the flow input to the flow-biasing network can be made during the functional tect by direct meter reading (IEEE 279 Standard for Nuclear Power Plant September 13, 1966).

Pr to ti n Systems, Section 4.9, O There are several instruments which must be calibrated, and will take several days to perform the calibration of the entire network. While the calibration is being performed, a zero flow signal will be sent to half of the APRMs, resulting in a half scram and rod block condition.

Thus, if the calibrations were performed during g operation, flux shaping would not be possible. Based on experience at other generating stations, drift of instruments such as those in the flow-biasing network, is not significant; thereforo, to avoid spurious scrams, a calibration frequency of each refueling outage is established.

O Reactor low water level instruments 1(2)-263-57A, 1(2)-263-57B, 1(2)-263-58A, and 1(2)-263-58B have been modified to be an analog trip system. The analog trip s system consists of an analog sensor (transmitter) and a master / slave trip unit setup which ultimately drives a U trip relay. The frequency of calibration for the trip unit has been established in General Electric topical Report NEDC-30851P-A as quarterly. An adequate for the calibration / surveillance test interval transmitter is once per operating cycle.

O Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with NEDC-30851P-A, " Technical Specification Improvement Analysis for BWR Reactor Protection System,"

as approved by the NRC in a letter dated July 15, 1987 from A. Thadani to T.A. Pickens .

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QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 O

The turbine control valve fast acting solenoid valve pressure switches directly measure the trip oil pressure that causes the turbine control valves to close in a rapid manner. The reactor scram setpoint was developed in accordance with NEDC-31336 " General Electric O Instrument Setpoint Methodology" dated October,1986. As part of the calculati.on, a calibration period is inputted to achieve a nominal trip point and an allowable setpoint (Technical specification value) . The nominal setpoint is procedurally controlled. Based on the calculation input, the calibration period is defined to be every Refueling O outage. .

The sensitivity of LPRM detectors decreases with exposure to neutron flux at a slow and approximately constant rate. Changes in power distribution and electronic drif t also require compensation. This compensation is O accomplished by calibrating the APRM system overy 7 days using heat balance data and by calibrating individual LPRM's at least overy 1000 equivalent full-power hours using TIP traverse data. Calibration on this frequency assures plant operation at or below thermal limits.

O A comparison of Tables 4.1-1 and 4.1-2 indicates that some instrument channels have not been included in the latter table. These are mode switch in shutdown, manual scram, main steamline isolation valve closure, and turbine stop valve closure. All of the devices or sensors associated with these scram functions are simple on-off switches, hence calibration la not applicable, o i.e., the switch is either on or off. Further, these switches are mounted solidly to the device and have a ,

very low probability of movingt e.g., the thermal switches in the scram discharge volume tank. Based on the above, no calibration is required for these g instrument channels.

B. The MFLPD.shall be checked once per day to determine if the APRM scram requires adjustment. This may normally be done by checking the LPRM readings, TIP traces, or process computer calculations. Only a small number of g control rods are moved daily, thus the peaking factors are not expected to change significantly and a daily check of the MFLPD is adequate.

O B 3.1/4.1-7 O ,

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QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 References

1. Licensing Topical Report NEDO-21617-A (December 1978).

> 2. General Electric Topical Report NEDC-30851P-A.

3. NEDC-31336 " General Electric Instrument Setpoint Methodology" dated october, 1986.

B 3.1/4.1-8

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O EXISTING TECH SPEC O

TS 3.1/4.1

'O ' REACTOR PROTECTION SYSTEM' O

O O .

o O

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l O L QUAD-CITlo OPR 29 3.1/4.1 REACTOR PROTECTION SYSTEM O

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 1

Applicability: Applicability: l Applies to instrumentation and associated Applies to the surveillance of the devices which initiate a reactor scram, instrumentation and associated devices O which initiate reactor scram.

Objective: Objective:

To assure the operability of the reactor To specify the type and frequency of protection system. surveillance to be applied to the O protection instrumentation.

SPECIFICATIONS A. The setpoints, minimum number of trip A. Instrumentation systems shall be systems, and minimum number of functionally tested and calibrated as O instrument channels that must be indicated in Tables 4.1-1 and 4.'-2 operable for each position of the respectively, reactor mode switch shall be as given in Tables 3.1-1 through 3.1-4. The

-system response times from the

(,pening of the sensor contact up to and including the opening of the trip O actuator contacts shall not exceed 50 milliseconds.

B. If, during operation, the maximum B. Daily during reactor power operation, fraction of limiting power density the core power distribution shall be

, exceeds the fraction of rated power checked for maximum fraction of O when operating above 25% rated limiting power density (HFLPD) and thermal power, either: compared with the fraction of rated power (FRP) when operating above 25%

rated thermal power.

1. The APRM scram and rod block settings shall be reduced to the O values given by the equations in Specification 2.1.A.1 and 2.1.B. This may also be accomplished by increasing the APRM gain as described therein, i

P 3 1/4 1*1 ^*'"d* * "t " 12 4

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QUAD-CITIES DPR 29

2. The power distribJtion shall be changed such tF.s the maximum fraction of l'.miting power density no *.onger exceeds the fraction of rated power.

C. When it is determined that~ a chant is f ailed in the unsafe cond)..on and Column 1 of Tables 3.1 1 through 3.1-3 cannot be met, that trip system must be put in the tripped condition immediately. All other RPS

. channtis that monitor the same variable sh&ll be functionelly tested within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The trip rystem with the f ailed enannel may be untrippejl,for a period of time n C to exceed I hour to conduct this testing. As long as the trip system with the f ailed channel contains at least one operable channel monitoring that same variable, that trip system may be placed in the untripped position for short periods of time to allev functional testing of all RPS instrument channels as spicified -

by Table 4.1-1. The trip system may be in the untripped position for no more than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> per functional test period for this testing.

3.1/4.1-? Amendment No. 114

_ _ _ _ _ _ _ _ - . _ _ _ _ _ __ _ ___ . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ . _ _ _ _ _ _ _ - __ -_- ___ . . J

g .

O QUAD-CITIES DPR-29 0 3.1 LIMITING CONDITIONS FOR OPERATION BASES The reactor protection system automatically initiates a reactor scram to:

a. preservetheintegrityofthefuelcladding g b. preserve the integrity of the primary system, and v

c. minimize the energy which must be absorbed and prevent criticality following a loss-of-coolant accident.

This specification provides the I/dkiMf//dllY//8/4/ $7/at///f necessary to preserve the ability of the system to tolerate single failurcs and still perform its intended

~

n functiorffpven dut ing periods when instrument channels may be outeofEtervice because of maintena'fice. When necessary, one channel may be made inoperable for brief intervals to conduct required functional tests and calibrations.

The reactor protection system is of the dual channel type (reference SAR Section 7.7.1.2$Thesystemismadeupoftwoindependenttripsystems,eachhavingtwo subchannels of tripping devices. Each subchannel has an input from at least one O instrument channel which monitors a critical parameter.

The outputs of the subchannels are combined in a one-out-of-tv gic i.e., an input signal on either one or both of the subchannels will cause a trip sys m trip. The outputs of the trip systems are arranged so that a trip on both systems is required to produce a reactor scram:

0--

This system meets Ahe requirements of the IEEE 279htandard for Nuclear Power Plant Protection SystemP' issued September 13, 1966. The systen. has a ' reliability greater than that of a two-out-of-three system and somewhat less than that of a one-out-of-two system (reference APED 5179).

at Icast With the exception of the average power ran monitor (APRH) and intermediate range o monitor (IRM) channels, each subchannel has one instrument channel. When the m'nimum

--condition for operation on the number of operable instrument channels per untripped protection trip system is met, or if it cannot be met and the affected protection trip

-**- system is placed in a tripped condition, the effectiveness of the protection system is preserved, i.e., the system can tolerate a single failure.and still perform its intended function of scramming the reactor. Three APRM instrument channels are provided for each protection trip tyttem.

O APRd# 1 and # 3 operate contacts in subchannelandAPRdf2and#3 operate contacts in the other subchannel. APR # 4, # 5 and # 6 are arranged similarly in the other protection trip system. Each protection trip system has one more APRM than is necessary to meet the rainimum number required per channel. This allows the bypassing of one APRM per protection trip system for maintenance, testing, or calibration.

Additional IRM channels have also been provided to allow for bypassing of one such O channel. The bases for the scram settings for the IRM, APRM, high reactor pressure, reactor low water level, turbine control valve fast closure, and turbine stop valve closure are discussed in Specifications 2.1 and 2.2.

O 3.1/4.1-3 Onendment No.114

4 s

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, QUAD-CITIES DPR-29 O Pressure sensing of the drywell is provided to detect a loss-of-coolant accident and initiate the emergency core cooling equipment. The pressure-sensing instrumentation is a backup to the water-level instrumentation which is discussed in Specification 2.1. A scram is provided at the same setting as the emergency core cooling system (ECCS) initiation to minimize the energy which must be accommodated during a loss-of-coolant accident and to prevent the reactor from going critical following the accident.

' The control rod drive scram system is designed so that all of the water which is discharged from the)(eactor by a scram can be accommodated in the discharge piping. A part of this system 7s an individual instrument volume for each of the south and north CRD accumulato These two volumes and their piping can hold in excess of 90 gallons of water and, e low point in the piping. No credit was taken for these volumes in the design of the discharge piping relative to the amount of water which must be o accommodated during a scram. During normal operations, the discharge volumes are empty; however, should either volume fill with water, the water discharged to the piping from the /eactor may not be accommodated which could result in slow scram times or partial or no control rod insertion. To preclude this occurrence, level switches have been

'nstalled in both volumes which will alarm and scram the /eactor when the volume t emaining in either instrument volume is approximately 40 gallons. For diversity of level sensing methods that will ensure and provide a scram, both differential pressure J switches and thermal switches have been incorporated into the design and logic of the system. The setpoint for the scram signal has been chosen on the basis of providing sufficient volume remaining to accommodate a strannyven with 5 gpm leakage per drive into the SDV. As indicated above, there is suffictient volume in the piping to accommodate the scram without impairment of the scram times or the amount of insertion of the control rods. This function shuts the /eactor down while sufficient volume 3

remains to accommodate the discharged water and precludes the situation in which a scrar

" would be required but not be able to perform its function properly. .

Loss of condenser vacuum occurs when the condenser can no longer handle heat input.

Loss of condenser vacuum initiates a closure of the turbine stop valves and turbine bypass valvegtyhich eliminates the heat input to the condenser. Closure of the turbine stop and bypaYs valves causes a pressure transient, neutron flux rise, and an increase J

in surface heat flux. To prevent the cladding safety limit from being exceeded if this occurs, a reactor scram occurs on turbine stop valve closure. The turbine stop valve closure scram function alone is adequatr. to prevent the cladding safety limit from being

- exceeded in the event of a turbine trip transient with bypass closure.

The condenser low-vacuum scram is a backup to the stop valve closure scram and causes a scram before the stop valves are closed, thus the resulting transient is less severe.

> Scram occurs at 2 Einches Hg vacuum, stop valve closure occurs at 2 O nches Hg vacuum, and bypass closure at Binches Hg vacuum.

O 3.1/4.1-4 Amendment No. Ill j

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k QUAD-CITIES DPR-29 D

High radiation levels in the main steamline tunnel above that due to the normal nitroge and oxygen radioactivity are an indication of leaking fuel. A scram is initiated whenever such radiation level exceeds fif teen times nomal background (without hydrogen addition). The purpose of this scram is to reduce the source of such radiation to the extent necessary to prevent excessive turbine contamination. Discharge of excessive amounts of radioactivity to the site environs is prevented by the air e.jector of f gas B monitors, which cause an isolation of the main condenser off-gas line provided the limit specified in Specification 3.8 is exceeded.

The main steamline isolation valve closure scram is set to scram when the isclation w ives are 10% closed from full open. This scram anticipatis the pressure and flux transient which would occur when the valves close. By scramming at this setting, the resultant transient is insignificant.

D '

A reactor mode switch is provided which actuates or bypasses the various scram functions appropriate to the ptrticular plant operating status (reference SAR Section 7.7.1.2).

Whenever the reactor mode switch is in the Rff% or S$/f#/Htf SfM4f position, the turbine condenser low vacuum scram and main steamline isolation valve closure scram are bypassed. This bypass has been provided for flexibility during startup and to allow D repairs to be made to the turbine condenser. While this bypass is in effect, protectior is provided against pressure or flux increases by the high pressure scram and APRM 15%

scram, respectively, which are effective in tW-mode: .sraerve/uor armoSY-m +qken If the reactor ere / ught to a hot standby condition for rapairs to the turbine condenser, the main steamline isolation valves would be closed. No hypothesized single failure or single operator action in this mode of operation can result in an unreviewed D radiological release. _

,e m n w n w oes The manual scram function is active in all+* odes, thus providing for a manual means of rapidly inserting control rods during all m;d;; cf reactor eptr;thn.orsanoNat moon.

The IRM system provides protection against excessive power levels and short reactor periods in the startup and intermediate power ranges (reference SAR Sections 7.4.4.2 and E 7.4.4.3). A-sevece range = niter (SRM) syst= S che pr:vid:d to-: pply addtt4enal laserf 'N -eevtren hv:1 hformat4:n durkg ;tertup 5.:t ha: n; ;;r= Nncthn: (refeeenc+-sag -

6ection 7.4.3.2). Thua-th; IRM h r: quired h the-M/#7A"nd+k-S!!/////HM-G#f;47 effE in eddition, p*etectica is provided ia + m -

• D ou m' "- "

disc +ned i- the4ase hr Specif ket4:n 2.1. In the power range, the APRM system provides required protecti reference SAR ,Section 7.4.5.2). Thus, the IRM system is y not required in the RM // the APRfFs cover only the intermediate and power range; the IRM T provide adeq ta c verage in the startup and intermediate range.

opezaroc o t h The higdeactor pressure, higN9d'rywell d pressure,Mwater level y and-+pam-dischaeg: vol = high hvel :cr=; u. Weed-fer th; Start /l st Stendby ond Nn eedes of-plant--eperat4cm They are therefore riaired to be operational for ese f //# of reactor operation, cessa rios at.

D

'~3unms art required. for opgearroNAL mooES t ernd. L . The scrurn discharge volumc high level scram is regw' red in DPGRarIDHAL.

InobES Is t And S  ;

1 D 3.1/4.1-5 Amendment No. 114

I I

5 D

QUAD-CITIES OPR-29 I The turbine condenser low-vacuum scram is required only during power operation and must be bypassed to start up the unit.

The require:ent--th:t4he-4M4--be in rted-4n-the-cer: when th: ^ " P", ' r :d 3/126-of M4-teele s;50res thst there is preper 0,eriop -;n the neuii un muni turiuh =y ivm> and ttiu. Uiai adryuoic sv1riege is previded for eli rer.ge; cf reeeter cperation.

D D

D D

3.1/4.1-6 Aalendment No. 114

o. ,

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l O

QUAD-CITIES DPR-29 0 4.1 SURVEILLANCE REQUIREMENTS BASES A. e minimum functional testing frequency used in this specification is based a liability analysis using the concepts developed in Reference 1. This the conc t was specifically adapted to the ene-out of-two taken twice logic reacto rotection ystem. The analysis shows that the sensors are pri rily resp nst e f r the reliability of the reactor protection system.. Th -

O analysis m es use of " unsafe failure" rate experience at conventio 1 and nuclear powe plants in a reliability snodel for the system. An u e afe to failure its is defined as e which negates channel operability and which, nature, is revea ed only when the channel is functionally tes d or attempts to respond to a real'N gnal. Failures such as blown fuses, ru ured bourdon tubes, f aulted amplu ers, f aulted cables, etc. , which re t in " upscale" or "downstale" readings o the reactor instrumentation are safe" and will be O easily recognized by the perators during operation b ause they are revealed by an alarm or a scram.

The channels listed in Table . -1 and 4.1-2 ar divided into three groups respecting functional testing.

O ,

These are:

1. On-off sensors that provide a ser rip function (Group 1);

2. Analog devices coupled with bi able tr 5 that provide a scram function (Group 2);- and,

3. Devices which serve a us ul function only du ng some restricted mode of operation, such as St up/ Hot Standby, Refuel, r Shutdown, or for which the only practical t it is one that can be perfor d at shutdown (Group 3).

The sensors that ma up Group 1 are specifically selecte from among the whole family of industri on-off sensors that have earned an exc lent reputation for reliable ope tion. Actual history on this class of sens s operating in O nuclear power ants shows four failures in 472 sensor years, oAa failure rate of 0.97 X 10 /hr. During design a goal of 0.99999 probability oT success (at the 50% co idence level) we. adopted to assure that a balanced an dequate of the design i achieved. The probability of soccess is primarily a funct sensor ailure rate and the test interval. A 3-month test interval was lanned for cup 1 sensors. This is in keeping with good operating practice an sa sfies the design goal for the logic configuration utilized in the react r m

V otection system.

Insert "8" O

3.1/4.1-7 Amendment No. 114 '

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0 O

QUAD-CITIES OPR-29 O

T satisfy the long-term objective of maintaining an adequate level of safpty thro hout the plant lifetime, a minimum goal of 0,9999 at the 95% conJJtience level reposed. With the one-out-of-two taken twice logic, thjptequires that each hosor have an availability of 0.993 at the 95% con nce level.

This level of 1,1 ability may be maintained by adjustin test interval as a function of the served failure history (Reference D. To facilitate the O implementation of th technique, Figure 4.1-1_is yra1 Tided to indicate an appropriate, trend in tes interval. The proce@r"e is as follows:

1. Like sensors are pooled in one gr for the purpose of data acquisition.

2. _The factor M is the exposure and is equal to the number of sensors in a group, n, times the el d time nT).

3. -The accumulated nu of unsafe failure is plotted as an ordinate against M as an absciss n Figure 4.1-1.

4. After a t d is established, the appropriate mo ly telt interval to sat he goal will be the test interval to the of the plotted s.

.O A test interval of 1 month will be used initially until a t d is established. -

r-movs To l The turbine control valve fast acting solenoid valve pressure switches directly arrst measure the trip oil pressure that causes the turbine control valves to close in a LAsr rapid manner. The reactor scram setpoint was developed in accordance with NEDC n enmapu 31336 " General Electric Instrument Setpoint Methodology" dated October, 1986. As

'd oN Ncyr part of the calculation, a calibration period is inputted to achieve a nominal trip PAqs point and an allowable setpoint (Technical Specification value). The nominal set

• point is procedurally controlled. Based on the calculation input, the calibration period is defined to be every Refueling Outage. ._

2 devices utilize an analog sensor followed by an amplifier and a hist e tri uit. The sensor and amplifier are active components, and a fai is n

v almost ways accompanied by an alarm and an indication of the source trouble.

In the even failure, repair or substitution can start immedia y. An as-is failure is one " sticks" midscale and is not capable of golffg either up or down in response to an ou f-limits input. This type of fail for analog devices is a rare occurrence and is ctable by an operator who o rves that one signal does not track the other three. purposes of analys , it is assumed that this rare

,g failure will be detected within curs.

The bistable trip circuit which is a p the Group 2 devices can sustain unsafe failures which are revealed only on . erefore, it is necessary to test them periodically.

A study was conducted of th nstrumentation channe neluded in the Group 2 devices to calculate ' unsafe' failure rates. The log devices (senso y and d, ampli?' m ) are pr d to have an unsafe failure rate o s than 20 X 10 failures / hour. bistagletripcircuitsarepredictedtohav unsafe failure rate of les an 2 X 10 failures / hours. Considering the 0-hour itoring interva the analog devices as assumed above and a weekly test inte 1 for the bist trip circuits, the design reliability goal of 0.99999 is attained h ample margin. l l

O 3.1/4.1-8 Amendment No.129 l

O-O I

quad-CITIES DPR-29 O The-tdetebc.c eevices :re : enitere4_4,an,-p;.nt_.perat,5.n_te_ rec- r4_their4 52ure history-end-ettetMsh-e-test 4*tervel-ttsing-the curve ef rip e 4-1-1-There are numerous--4denthal- b!: table-dwices_used-.-throughout-the-plant-instreientet4on syrterThei riwi e, eignif4 cent-dat+-on-the f ailure ratee far tha hittable deviges shoulet ha are"mniatad ranidly.

Qarter}

The frequency f calibration of the APRM flo Ebiasing network has been established r) at each refu ing outage. The flo4 biasing network is functionally tested at least once per ,and, in addition, cross calibration checks of the flow input to the flow-biasing network can be made during the functional test by direct meter reading (IEEE 279 Standard -for Nuclear Power Plant Protection Systems, Section 4.9, September 13 1966). There are several instruments which must be calibrated, and it will take several days to perform the calibration of the entire network. While the O cali tion is being performed, a zero flow signal will be sent to half of the AP esulting in a half scram and rod block condition. Thus, if the calibratior@

were formed during operation, flux shaping would not be possible. Based on experience at other generating stations, drift of instrumen@,such as those in the flo4)iasing networgis not significant; therefore, to avoid spurious scrams, a calibration frequency of each refueling outage is established.

O Reactor low water level instruments 1-263-57A, l'-263-57B,1"h63-5BA, and 1-$63-58B have been modified to be an analog trip system. The analog trip system consists of an analog sensor (transmitter) and a master / slave trip unit setup which ultimately drives a trip relay. The frequency of calibration and function;l test 4*g4er h=t instrument 100pc Of the anale; trip syster, including r4.ctor law W er leue1 beea esteblished ia Licensing Tepica' % pert NE00-21617 ^ (December 1978). M th the ene-cut of-tuc-t: hen-tw4cc 10gic, NED0416W-A-stetes-that esch trip widt-tre o#

subjceted to e celibration/ functional tett of enc month. An adequate calibration / surveillance test interval for the transmitter is orice per operating cycle.

Qce.ified surveillance intervals and. surveillance and. maihlengI1x g ca.fage fimcs have been determineel in acconfar1cc. wilh NEDC -308 51f-A,

• Technica I Apecificalicr) Improvemen+ Analyses for bulR Reacdor Pre f ec fior) S s te m " , a.s aff re v ed. but the NRL in a lellCV*

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d a.tecL Ju.I is ,199 4 from A. Th ada n i to T A , Pickeru

[ Acid paragraph frorn page s I/ a, t -a ]

() for +he trip unit ha s been <siablished in l]eneral Glec}r?

Topical Re,porf NEDC: 30851P-A as quarf erly.

O 3.1/4.1-9 Amendment No. 114 O

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O 4 QUAD-CITIES DPR-29 w,

p 3 devices are active only during a given portion of the operati . For 1 the IRM is active during startup end inactive during fugil-wer operation.

Thus, th ly test that is meaningful is the one performed ju t prior to shutdown or startup, . the tests that are performed just prior use of the instrument.

O Calibration frequency the instrument channel is ided into two groups. These are as follows:

1. Passive type indicating device aFcan be compared with like units on a continuous basis, and

, 2. Vacuum tube or semicon or devices and de - ors that drift or bse J sensitivity.

Experience with ssive type instruments in Commonwealth Ediso nerating stations and substat indicate that the specified calibrations are adequa For these devices ch employ amplifiers, etc. drift specifications call for dr 4obelest .

tha .4%/ month 1.e. , in the period of a month a drif t of 0.4% would occur, Itg viding for adequate margin.

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The sensitivity of LPRM detectors decreases with exposure to neutron flux at a slow and approximately constant rate. Changes in a power distribution ar.d electronic drift also require compensation. This compensation is accomplished by calibrating the APRM system every 7 days using heat balance data by calibrating individual LPRM's at least every 1000 equivalent full power hours using TIP traverse data.

3 Calibration on this frequency assures plant operation at or below thermal limits.

A comparison of Tables 4.1-1 and 4.1-2 indicates that some instrument channels have not been included in the latter table. These are mode switch in shutdown, manual scram, high water level in scram discharge volume, main steamline isolation valve closure, and turbine stop valve closure. All of the devices or sensors associatedj with these scram functions are simple on-off switches, hence calibration is not g applicable i.e. , the switch is either on or off. Further, these swite.es are mounted solidly to the device and have a very low probability of moving; e.g., the thermal switches in the scram discharge volume tank. Based on the above, no cali-bration is required for these instrument channels.

B. The hFLPD shall be checked cnce per day to determine if the APRM scram requires adjustment. This may normally be done by checking the LPRM readings TIP traces, os g process computer calculations. Only a small number of control rods are moved daily, thus the peaking f actors are not expected to change significantly and a daily check of the MFLPD is adequate.

References

1. I. ".. Rech:, "". lkbility of En';ineceed hf:ty F::ture: a: : Fun %46n-M 4- Testir,g Frquer.cy", Lcleer S&fety, W i. 5, Hv. 4, pp. 310-312, July-Avgust 19683 1 7. Licensing Topical Report NEDO-21617-A (December 1978).

3. NEDC - 31336 " General Electric Instrument Setpoint Methodol'ogy" dated October, 1986 3 2. , GeneraI Dec1ric. Tcfi col Repor} HEOc .5D85IP-A 3.1/4.1-10 Amendment Ho.129

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INSERT FOR TECHNICAL SPECIFICATION -

& CTION 3.1/4.1" REACTOR PROTECTION SYSTEM" Insert "A" During refuelin g, the primary Neutron Monitoring System (NMS) indication

) of neutron flux levels is arovided by the Source Range Monitors (SRM). The >

SRMs provide input to the RPS, but shorting links are installed across the normally closed contacts such that tripping an SRM channel does not affect i

the RPS. To activate the SRM scram function, these shorting links must be removed from the RPS. The SRM control rod scram provides backup protection to refueling interlocks and SHUTDOWN MARGIN should a

')- prompt reactivity excursion occur. Although the IRM and APRM functions are req uired to be OPERABLE durin l

on-scale monitoring of neutron flux .5 refueling, the SRMs provid shorting links must be removed to enable the scram function of the SRMs.

The RPS (and therefore removal of the RPS shorting links) is required to be 3 - OPERABLE in REFUEL o with any control rod withdrawn from a core cell containing one or more fue assemblies. Control rods withdrawn from a core -

cell containing no fuel assemblies do not affect the reactivi of the core and therefore are not required to have the capability to scram. rovided all control rods are otherwise inserted, the RPS function is not required. In this condition, the required SHUTDOWN MARGIN and the one-rod-out interlock -

)' provide assurance knat the reactor will'not become critical thereb uiring'a scram. If the SHUTDOWN MARGIN has been demonstrated, th RF shorting links are not required to be removed. Under these conditione, the capability of the one-rod-out interlock to prevent critical has been demonstrated and the backup scram protection provided the IRMs is

)_ sufficient to ensure a highly reliable scram if required.

Insert "B" Su-veillance requirements for the reactor protection system are selected in order to demonstrate proper function and operability. The sauveillance

[

intervals aregood experience,2) determined different wa en ineerin in many'udgement,3)ys, such reliability analyses, or 4as,1) opera analyses that are foun accepta le to the NRC.

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O o SIGNIFICANT HAZARDS CONSIDERATIC?1S AND ENVIRONMENTAL ASSESSMENT EVALUATION O.

PROPOSED TS 3.1/4.1 O ' REACTOR PROTECTION SYSTEM'

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EVALUATION EQB BIGNIFICANT HAZARDS CONSIDERATION PROPOSED SPECIFICATION 3.1/4.1 REACTOR PROTECTION SYSTEM a

The proposed changes provided in this amendment request are made in order to provide a more user friendly document, incorporate desired technical improvements, and to incorporate some improvements from later operating BWRs. These changes have been reviewsd by Commonwealth Edison and we believe that they do not n present a Significant Hazards Consideration. The basis for our V determination is documented as follows:

BASIS IQB EQ SIGNIFICANT HAZARDS CONSIDEPATION Commonwealth Edison has evaluated this proposed amendment and determined that it involves no significant hazards O consideration. In accordance with the criteria of 10 CFR 50.92(c) a proposed amendment to an operating license involves no significant hazards consideration if operation of the facility, in accordance with the proposed amendment, would not:

1) Involve a significant increase in the probability or consequences of an accident previously evaluated, because:

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a. The Generic Changes to the technical specifications involve administrative changes to format and arrangement of the material. As such, these changes cannot involve a significant increase in the probability or consequences of an accident previously evaluated.

O b. The proposed changes to specifications 3.1.A/4.1.A and 3.1.B/4.1.B are made to provide the user with a format that will allow quicker access to needed information and to provide concise LCO, Applicability, Action and Surveillance requirements. The blend of requirements from the present Quad Cities Technical Specifications and later

() operating BWRs utilizes proven material and testing techniques. The deletion of Surveillance Re @irement 4.1.C on additional testing of RPS channels if one fails in the unsafe position does not significantly decrease the reliability of the RPS system. This additional testing may or may not find more problems in the system such as common mode failures. Evaluations to determine cause of O the failure and the potential for additional failures in similar equipment provides an equivalent level of safety in the plant as the present testing requirements of 4.1.C.

The proposed changes to Tables 3.1-1 through 3.1-3, 4.1-1 and 4.1-2 do not alter any established setpoints or reduce the minimum operable channels per trip system g' requirements. The propoced changes are applicable for the l

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O Quad Cities plant and are current plant operating practice or have been utilized on other operating plants; therefore, they do not involve a significant increase in the probability or consequences of an accident previously evaluated.

O c. The proposed changes to incorporate the Surveillance Testing Intervals and Allowed Out of Service Intervals in Topical Report NEDC-30851P-A do not degrade the reliability of the RPS system, as demonstrated in the Topical Report and corresponding plant specific analyses.

Section 5.7.4 of NEDC-30851P-A provides a detailed generic O Determination of No Significant Hazards for the proposed change. Implementation of the extended surveillance intervals Calibrations the for will Channel Functional not be made without Tests and Channel factoring in appropriate drift information into the setpoint calculations. Since the changes do not degrade the reliability of the RPS system over present conditions, O there is no significant increase in the probability or consequences of an accident previously evaluated,

d. The proposed change to delete the APRM Downscale Scram Trip Function has been evaluated by Commonwealth Edison and General Electric. The accidents of concern with c respect to the APRM/IRM companion trip are the Rod Drop d Accident (RDA) and the low power Rod Withdrawal Error (RWE). FSAR and reload safety analyses do not credit this scram function in the termination of either of these accidents. Since this scram function is not credited in the termination of these accidents, the elimination of this scram function has no adverse effect of previously O evaluated accidents.

2) Create'the possibility of a new or different kind of accident from any previously ev aluated because:

a. Since the Generic Changes proposed to the technical specifications are administrative in nature, they cannot n

s create the possibility of a new or different kind of accident from any previously evaluated.

b. The changes to Specifications 3.1.A/4.1.A and 3.1.B/4.1.B blend STS raquirements with existing Quad Cities requirements to provide a user friendly format and g -presentation of requirements. The deletion of

' Surveillance Requirement 4.1.C concerning additional testing of RPS channels if a channel fails in an unsafe position does not create the potential of a new or ,

different kind of accident since other means are utilized to determine potential for common cause or similar failures in other channels.- Many of the changes proposed

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to the Tables follow later operatirg BWR guidelines that are presently being utilized at these plants and have been l'

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O evaluated and found acceptable for use at Quad cities.

Other changes to the tables provide clarification of present requirements. Therefore, the changes do not create the possibility of a new or different kind of accident from any previously evaluated.

O c. The proposed changes to incorporate the Surveillance Testing Intervals and Allowed out of Service Times in Topical Report NEDC-30851P-A do not create the possibility of a new or different kind of accident from any previously evaluated because RPS function and reliability is not degraded

• y these changes. No new modas of plant O operation are involved. The implementation of STS Channel Calibration Test frequencies will only be made to the extent that the instrumentation drift characteristics allow the interval extensions.

d. The deletion of the APRM Downscale Scram Trip Function does not introduce any new accident scenario. The U limiting accidents (i.e., RDA and RWE) in the operating region of transition between the Startup and Run Operational Modes are well understood and are evaluated in FSAR and/or reload safety analyses. Other control rod initiated events which are less limiting in this region, such as fast period events (either due to operator error n or CRD malfunction), are subsets of the low power RWE

" event and are bounded by it and the Design Basis RDA.

General Electric has indicated that, for reactivity insertion mechanisms at very low power (if postulated to occur coincident with an inappropriate. mode switch), the only effect of the deletion of the APRM downscale scram would be that tha initial power level could be a few O percent lower which would not have a significant effect on the severity of the event. In addition, proper overlap between the 1RMs and APRMs is not affected since the calibration requirements cre not being changed.

3) Involve a significant reduction in the margin of safety because

,a

a. The Generic Changes proposed in this amendment request are administrative in nature and, as such, do not involve a reduction in the margin of safety.

b. The changes to Specifications 3.1.A/4.1.A and 3.1.B/4.1.B n implement.an STS type of format while retaining the

present two column layout. This two column layout has been in use at Quad Cities since initial licensing and is preferred by the majority of the technical specification users at the plant. The proposed LCo, Applicability, Actions and Surveillance Requirements are modeled after STS requirements which have bean evaluated and found to be O acceptable for use at Quad Cities. The deletion of present Surveillance Requirement 4.1.C does not involve a

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O significant reduction in the margin of safety since other equivalent methods are utilized to determine if the failure of one RPS channel in an unsafe position affects other similar channels.

The changes to the Tables in Section 3.1/4.1 follow proven

.O STS guidelines that have been implemented at other operating BWR plants. These changes have been evaluated for use at Quad Cities with a determination that implementation at the plant will not involve a significant reduc' ion in the margin of safety. Other changes to the tr.bles involve clarifications or minor improvements that 47 not affect the margin of safety.

g

c. Tha changes proposed in Topical Report NEDC-30851P-A increases the testing for the Manual Scram function and decreases testing for the other applicable Scram functions. Allowed out of service times are increased for the RPS channels as a result of the Topical Report 9' analyses. However, the requested changes do not degrade the reliability of the RPS system and thus the margin of safety is preserved. The results of the topical report have been found acceptable for plant use by NRC SER with the stipulation that setpoint drift over the increased testing interval be considered in setpoint calculations.

Quad Cities will consider the additional drift in the O setpoint calculations before implementing the extended surveillance testing intervals for both the Channel l

Punctional Tests and the Channel Calibration Tests.

i the changes do not involve a significant Therefore,in reduction the margin of safety.

, d. The APRM Downscale Scram Trip Function is not credited in U the termination of any FSAR or reload safety analysis l event. As such, the elimination of this scram function has no effect on any margin of safety, l

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ENVIRONMENTAL ARJESSMENT EVALUATION PROPOSED SPECIFICATION C). SECTION 3.1/4.1 REACTOR PROTECTION SYSTEM Commonwealth Edison has evaluated the proposed amendment in accordance with the requirements of 10 CFR 51.21 and has determined that the amendment meets the requirements for categorical exclusion as specified by 10 CFR 51.22 (c) (9) .

() Commonwealth Edison has determined that the amendment involves no significant hazards consideration, there are no significant change in the types or significant increase in the amounts of any effluent that may be released offsite, and there is no significant increase in individual or cumulative occupational radiation exposure.

O The proposed amendment does not modify the existing facility and does not involve any new operation of the plant.

As such, the proposed amendment does not involve any change in the type or significant increases in effluents, or increase individual or cumulative occupational radiation

() exposure. The proposed amendment to Section 3.1/4.1,

" Reactor Protection System" contains administrative changes such as including appropriate applicability statements within the specifications to define the applicability during operating mode and the required actions to be implemented in the event that specification cannot be met. The added O requirements are based on Standard Technical Specifications and later operating plant requirements. The proposed specification also arranges the tables to provide for user-friendly presentation.

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3 QC-1/ QC-2 DIFFERENCES 01 .

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COMPARISON OF UNIT 1 AND UNIT 2 TECHNICAL SPECIFICATIONS FOR THE IDENTIFICATION OF TECHNICAL DIFFERENCES O SECTION 3.1/4.1 REACTOR PROTECTION SYSTEM Commonwealth Edison has conducted a comparison review of C) the Unit 1 and Unit 2 Technical Specifications to identify any technical differerces in support of combining the Technical Specifications into one document. The intent of the review was not to identify any differences in presentation style (e.g. table formats, use of capital letters, etc.), punctuation, or spelling errors but rather to 0 identify areas which the Technical Specifications are technically or administrative 1y different.

The review of Section 3.1/4.1 " Reactor Protection System" revealed the following technical differences:

O Note (8) of " Notes for Tables 3.1-1, 3.1-2 and 3.1-3" (Page 3.1/4.1-14 for DPR-29) contains the statement, "1 inch on the water level instrumentation is > 504" above vessel zero (See Reference Bases 3.2)." which is not contained in the Unit 2 Technical Specification. This information is accurate for application on both units. Unit 1 and Unit 2 O Technical Specification Bases section 3.2 contains the background for this statement.

Several administrative differences were identified as follows:

Page 3.1/4,1-1 O

Applicability Unit 1: Applies to instrumentation and ...

Unit 2: Applies to the instrumentation and...

Page 3.1/4.1-3 O Unit 1: minimize the energy which must be Paragraph 1 c.

absorbed ...

Unit 2: c. minimize the energy which must be adsorbed ...

Pace 3.1/4.1-4 O

Paragraph 2 Unit 1: into the SDV. As indicated above, there...

Unit 2: into SDV. As indicated above, there...

O Paragraph 3 Unit 1: Loss of condensor vacuum...

Unit 2: Loss of condensate vacuum...

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• Pace 3.1/4.1-5 O

Paragraph 7 Unit 1: discharge volume high level scrams are required for the Start / Hot Standby...

Unit 2: discharge volume high level scrams are required for the Startup/ Hot

() Standby...

Pace 3.1/4.1-7 Paragraph 2 Unit 1: The channels listed in Table ....

Unit 2: The channels listed in Tables ...

O Page 3.1/4.1-8 Paragraph 2 Unit 1: rare occurrence and is detectable by an operator who observes that one signal...

rare occurrence and is detectable by O Unit 2:

an operator who observes than on signal.... ,

Paragraph 4 Unit 1: amplifiers) are predicted to have ...

Unit 2: amplifiers) are predicated to have..

O failures / hour. The bistable trip Paragraph 4 Unit 1:

circuits are predicted to have ...

Unit 2: failures / hour. The bistable trip circuits are predicated to have ...

O Paragraph 4 Unit 1: rate of less than 2 x 10 (-6) failures / hours Unit 2: rate of less than 2 X 10 (-6) failures / hour ...

Paragraph 4 Unit 1: bistable trip circuits, the design reliability goal of 0.99999 is O attained with ...

Unit 2: bistable trip circuits, the design reliability goal of 0.99999 is attained wity ...

Paracraoh 3.1/4.1-9 O

Paragraph 2 Unit 1: once per month and, in addition, cross calibration checks of the flow

-Unit 2: once per month and, in addition, cross calibration check of flow ...

O Paragraph 3 Unit 1: subjected to a calibration / functional test of one month ...

Unit 2: subjected to a calibration / functional I test frequency of one month ...

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o Pace 3.1/4.1-10 Paragraph 1 Unit 1

' Group 3 devices are active only during-a given portion of the l operation. cycle ...

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Unit 2

aroup 3 devices are active only

-during a-given portion.of the operational cycle 1...;

Paragraph 3 Unit 1: and-substations indicate that Unit 2: and substations indicates that

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= Paragraph-4 Unit-1. and approximately constant rate.

Changes in a power' distribution ...

Unit 2: and approximately constant rate.

Changes in power. distribution ...

N)-- Paragraph 4 Unit 1: .the-APRM system every 7 days using heat balance data by calibrating ...

Unit 2: . the APRM; system every 7. days using heat balance data and by calibrating Pace 3.1/4.1-16

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-Note-[9] Unit 1: electronic calibrator during the three month test ...

Unit 2:- electronic-calibrator during the three month interval test ...

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