ML20076D880

Proposed Tech Specs Section 3.1/4.1, Reactor Protection Sys

ML20076D880
Person / Time
Site:	Quad Cities
Issue date:	07/23/1991
From:	COMMONWEALTH EDISON CO.
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ML20076D878	List: ML20076D877 ML20076D880
References
NUDOCS 9107300302
Download: ML20076D880 (45)

Text

.- . .. .

& a PROPOSED TECH SPEC TS 3.1/4.1

' REACTOR PROTECTION SYSTEM' l

9107300302 910723 PDR ADOCK 05000254 p PDR

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QUAD CITIES UNITS 1& 2 DPR-29 & DPR-30 3.1/4.1 REACTOR PROTECTION SYSTEM SPECIFICATIONS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 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:

and minimum number of instrument CHANNELS 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 up to and including the opening tested, calibrated and of the trip actuator contacts checked as indicated in shall not exceed 50 milliseconds. Tables 4.1-1 and 4.1-:.

hPPLICABILITY: 2. The system responne times for each Trip Function shown l- As shown .in Table 3.1-1. in Table 3.1-1 shall be l demonstrated to be within I ACTION: its limit at least each REFUELING OUTAGE. Each test

1. With a reactor protection shall include at least one j system instrumentation CHANNEL per TRIP SYSTEM such

setpoint less conservative that all CHANNELS are tested l than the value shown in the at least once every (N) l Trip Level Setting column of REFUELING OUTAGES where (N)

Table 3.1-1, declare the is the total number of CHANNEL inoperable and redundant CHANNELS in a follow ACTION 3.1.A.2 or specific reactor TRIP.

3.1.A.3 below until the SYSTEM.

CdANNEL is restored to OPERABLE status with its setpoint adjusted consistent with the Trip Level Setting value.

! 2~. With the number of OPERABLE l CHANNELS less than required l by the Minimum OPEPABLE l CHANNELS per TRIP SYSTEM requirement for one TRIP SYSTEM, place tne inoperable.

CHANNEL (s) and/or that TRIP SYSTEM in the tripped 3.1/4.1-1 l

v 4 QUAD CITIES UNITS 1 &2 DPR-29 & DPR-30 condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

An inoperablo CHANNEL nood not be p13ced in the tripped  ;

condition when this would

• cause the PROTECTIVE FUNCTIC11 to occur. In thoso casos, the inoporable CilANNEL shall be rostored 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 firct dctormined to be inoperablo or the ACTION required by Table 3.1-1 shall be ontored.

3. With the number of OPERABLE CllANNELS lobs than required by the Minimum OPERABLE CilANNELS per TRIP SYSTEM requiremont 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 then take the ACTION requiled by Table 3.1-1.

The TRIP EYSTEM need not be placed in the tripped condition if this would cause the PROTECTIVF [

FUNCTION to occur. When a TRIP SYSTEM can be placed in the tripped condition without causing the PROTECTIVE FUNCTION to occur, place the TRIP SYSTEM with the mest inoperablo CHANNELS in the tripped condition; if both systems have the same number of inoperable CilANNELS , place either TRIP SYSTEM in the tripped condition.

B. APRM Scram and Control Rod Block D. APRM Scram and Control Rod Block Flow Biased Upscale Sotpoints Flow Biased Upscale Sotpoints The APRM flow biased noutron flux The core power distribution shall upscale scram trip sotpoint and be checked daily for MFLPD and flow biased neutron flux upscale compared with the FRP.

3.1/4.1-2 L

1

I y

• a QUAD CITIES UllITS 1 &2 DPR-29 & DPR-30 control rod block trip sotpoint shall be established according to l tho equations in Specifications 2.1.A.1 and 2.1.B.

, MELICADILITY t OPERATIollAL MODE 1, when thermal power is greator than or equal to 25% of RATED TilERMAL POWER.

ACTI0 fit  ;

1. With the APRM flow biased neutron flux upscale scram trip sotpuint and/or the flow biased neutron flux upscale control rod block trip sotpoint loss concorvativo than the valuo shown in the equations in Specifications 2.1.A.1 and 2.1.B, initiato correctivo action within 15 minutos and within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, adjust the sotpoints to be consistent with the Trip Sotpoint valuos or increase the APRM gain as described in Specification 2.1.A.1 and 2.1.B or reduco thermal power to less than 25% of RATED TflERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

3.1/4.1-3

O s QUAD CITIES UtilTS 1 & 2 DFR 29 & DPR.30 TABLE 3.1-1 PlACIOR_ PEOILC11Q!! SYSTD1_.11CEMLIMIECiGIAT101LEEn'lEMGIS Mintmum CPtPAstt Applicacle CHA!JNEL3 Per Trip Level CPE PATION AL TRIP SYSTEN (aitDi Setteng MODES ACTION Trip Tunesten i N.A. 1,  ; 1

1. Mode switch 3,

• 4 an $hutdown 1 5 2 1

i N.A. 1, 2 1

2. Manual scram

• 1 3 4 5 6 1

1

3. IRM (c) 4.-High Flux 3 1 120/125 of fuli 2 2 acale 3, 4 1 3snt $tm) 2  ;

b. Incperattve 1 N.A. 2 1 2 3, 4 1 Jtni 5 2 ,

4. APRM (f)

4. High rium 2- Tech S pc 2.1.A.1 1 3 (flow biased)

b. Inoperative 2 N.A. 1, 2 1

. 2 3 7 2(n) S(e) 2 .

c. High Flum 2 Tech Spec 2.1.A.2 2 1 3 7 (15% ocram) 2 2(n) 5(e) 2

d. High Flum 2 Tech Spec 2.1.A.1 1 3 (Scres clamp)

5. Reactor Nigh 2 1 1060 peig 1, 2(g) -- 1 Pressure Drywell Nigh 1 2.5 peig 1, 2(h) 1

6. 2 Preseure

7. Reactor Low 2 1 8 inches (d) 1, 2 1 Water Level 3.1/4.1-4 r

l

OUAD CITIES L'!1ITS 1 & 2 Di' A- 2 9 5 C I'R- 3 0 TABLE 3.1-1

'clhG R.IMIIC10iL1'LEII!L11CEAM L DEIP ?J M E '1 TAT 121LRE1'JJ EE ME!G S Minimum CPEPABLE Appiteacle CHAN NE LS Per Trap Levet OP E PAT I C RL 1 RIP SYSTEM (allb) Setting MODES ACTION Trtp runction s 40 gallons 1, 2 .

9. scram Disenarge 2/tann 2/bana 5(.,i.; .

Volune Hagn

.later Level

9. Turetne Condsneer 2 3 21 inenes Hg 1 4 Lew Vacuum vacuum s 15 X Normal Fuli 1, 2(g) 4

10. Main steam Line 2 High Radiatten Pcwor Background (e)

11. Main steam Line 4 14> s 10% Valve 1 3 Isolation valve closure C1osure

12. Turbine Control 2 3 160 psig (c) 1()) 5 Valve rast closure

13. Turbine stop 4 5 10% Valve 1(3) 5 Valve Closure C1coure

14. Turbine EHC 2 2 900 peig 1()) $

Control Fluid Low Pressure 3.1/4.1-5

v- e QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 TABLE 3.1-1 (Continued)

}{EACTOR PRQTEC1' ION _ SYSTEM (SCRAM) INSTRUMENTATION _ REOUIRE!dEllTS ACTIONS ACTION 1 - De in at least ilOT SilUTDOWN 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 insort all insortablo control rods within one hour.

ACTION 3 - Bo in at least STARTUP within a hours.

ACTION 4 - Bo in STARTUP with the main steam lino isolation valvos 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 ilOT SIIUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

ACTION 5 - Initiato a reduction in thermal power within 15 minutos and reduce turbino first stago pressure to that which corresponds to loss 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 <11 operations involving CORE ALTERATIONS *,

and incert all insertablo control rods and lock the reactor modo switch in the SHUTDOWN position within one hour.

ACTION 7 - Verify all insertable control rods to be inserted in the core and lock the reactor modo switch in the Shutdown position within one hour.

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

3.1/4.1-6

-~ .- .__ __ __ _ - _ _ _

V s QUAD CITIES UllITS 1 &2 DPR-29 & DPR-30 i TABLE 3.1-1 (Continued)

R E ACTO R PR OT E CT I O11_ S Y ST Eti_1S CR AM ) INSJRUME!JTATION_REOUIREME!JTS TABLE 110TATIONS (a) CitANNE13 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 required survoillance without placing the TRIP SYSTEM in the tripped condition provided at least one OPERABLE CilAN!1EL in the samo TRIP SYSTEM is monitoring that paramotor.

(b) Two TRIP SYSTEMS shall be OPERABLE in the applicablo OPERATIONAL '

MODES for the specified Trip runction. CilANNEL OPERABILITY requiromonts withLn the TRIP SYSTEM are specified in the ACTIOli provisions of Specification 3.1.A.

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

(d) The +8-inch trip point is the water level as measured by the instrumentation outside the chroud. The water lovel insido the shroud will decreano as power is increased to 100% in comparison to the levol outside the shroud to a maximum of 7 inches. This is due to the pressure drop across the steem dryor. Therefore, at 100%

power, an indication of +8 inch water lovel will actually bo +1 inch inside the shroud. 1 inch on the water level instrumentation is .>_

504" abovo vessel zero.

(e) CilAN!1EL shared by the reactor protection and containment isolation system.

(f) An APRM will be considered inoperable if there are fewer than 2 LPRM inputs per level or there are less than 50% of the normal complomont of LPRMs to an APRM.

(g) This function is not required to bo OPERABLE when the reactor pressure vossol head is not bolted to the vessel.

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

(i) With any control rod withdrawn.

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

Not applicable to control (j) Permissible to bypass when turbino first stago pressure is loss than

~

that which corresponds to 45% of rated steam flow (< 400 psi).

l 1

3.1/4.1-7 y rr --m'r '=u- ' - vi*w- - ww ----

= > - - - - - w 9'---m*+ - - - - *- ^

e e QUAD CITIES UllITS 1& 2  ;

DPR-29 & DPR-30 TABLE 3.1-1 (Continued)

BEACTOR. PROTICT1911 SYSTEliE.CEMil IllSTBUED{IATIOli REQt).IRGiffilTQ TABLE ff0TATlRRf1 (k) The design permits closure of any one line without a scram being )

initiated. 1 (1) Permissible to bypass, with control rod block, for reactor ,

protection system reset in REFUEL and SilVTDOWil positions of the reactor modo switch.

(m) The " shorting links" shall be removed from the RPS circuitry prior to and during the time any control rod is withdrawn and shutdown margin demonstrations are being performed. liot required for control rods removed por Specification 3.10.D or 3.10.E.

(n) The non-coincident !!!4s reactor trip function logic is such that all channels go to both trip systons. Thorofore, when the "chorting links" are removed, the 141nimum OPERABLE CllAltllELS Por TRIP SYSTEli is 4 APRiiS and 6 IR!4S .

(o) Trip is indicativo of turbino control valvo fast closure (due to low E!!C fluid pressuro) as a result of fast acting valvo actuation, i

{.

i 3.1/4.1-8

_ _ . ~ . - . - - - . - . - - -

. . . - . - - . - - . _ . . - - - . - ~ .

5 e o l QUAD CITIES UllI7S 1 & 2 a- DPR-29 & DPR-30 l l

TADLE 4.1-1

, 1 REACTOR PROTECTION SYSTEM (SCRAM) Iff STRUMENTATIOl{

CHANNEL FUNCTIONAL TEST AND CHANNEL C}jECK REOUIREMEllT3

CHANNEL CHANNEL Applicable FUNCTIONAL FUNCTIONAL CHANNEL C?ERATIONAL Trip Function TEST Method TEST CHECKS MODES J (a) (c)

1. Mode Switch in Place Mode Sw R NA 1, 2, 3, 4, 5

^

Shutdown in Shutdown 2, Manual Scram- Trip Channel W NA 1, 2 , 3, 4, 5- )

and Alarm j i

3. IRM

a. High Flux Trip Channel S/U(e), W S/U, S, (b) 2(k) ,

and Alarm (d) W S 3, 4, 5 i

b. Inoperative Trip channel W NA 2(k), 3, 4, 5 and Alarm

4. APRM

a. High Flux Trip output W(L), Q 3, D(1) 1 -

(flow biased) Relays (d) b '. Inoperative Trip output- Q NA 1, : 2 , 3, 5 Relays

c. HL'gh Flux Trip Output S/U(e), W S/U, S, (b) .2(k) *

(154 scram) Relays (d) W S 3, 5

d. High Flux Trip Output W(1), Q $ 1 l

~

(Scram Clamp) Relays (d)

5. Reactor High Trip Channel .Q .NA 1, 2(h)

Pressure and Alarm ,

-6. Drywell High-. Trip Channel Q NA 1, 2(j)

Pressure and Alarm ,

7. Reactor Low Trip Channel -Q D 1, 2 Water Level. and Alarm-3.1/4.1-9

I e e QUAD CITIES UllITS 1& 2 DPR-29 & DPR-30 TABLE 4.1-1 PEACTOE_2POTECTIO!! SYSTEM _fSCRAM) I.11 STRUM EL{I&ILQ1{

Cl{btillEL FU?iCTIOflAL TEST AtJD ClBill!EL_Cl@CK REOUIREME!JTji l CHANNEL Cr(ANNEL Applicable i FUllCTIONAL FUNCTIONAL CHANNEL OPEPATIONAL I Trip Function TEST Methods TEST CHECK MODES (a) (C)

8. Scram Discharge Trip Channel Q NA 1, 2, 5(g) volume High Wate- and Alarm (f)

Level (Thermal and dp Switch)

9. Turbine Con'anser Trip Channel Q NA 1 Low Vacuum and Alarm

10. Main Steam Line Trip Channel Q S 1, 2(h)

High Radiation and Alarm (d)

11. Main Steam Line Trip Channel Q NA 1 Isolation Valve and Alarm closure

12. Turbine control Trip Channel Q NA 1 valve ramt Closure and Alarm

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

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

3.1/4.1-10 l

i I

QUAD CITIES UllITS 1 & 2 DPR-29 & DPR-30 l TABLE 4.1-1 (Continued)

REACTOR EERT.gCTIOff SYSIEti_(HCRAM) IllSIIMiEllTAIlQll C11Al{1{EL FUtiCTIOllAL TEST AllD CilAlll{EL_CliEQLREOUIREtiEllIS IAllLE 110TATIO!!E (a) A CilAlfilEL FUtiCTIOliAL TEST of the logic of each C11Ati!1EL is performed as indicated. This coupled with placing the modo switch in Shutdown ,

each REFUELIliG OUTAGE cont.titutos a LOGIC SYSTEM FUllCTIO!1AL TEST of  !

the scram system. l (b) The IRM and SRM channels shall be datormined to overlap for at least (1/2) docados during each startup af ter entering OPERATIO!1AL MODE 2 ,

and the IRM and APRM channels shall be determined to overlap for at least (1/2) decados during each controlled shutdown, if not performed within the previous 7 days.

(c) CilAllt1EL FUllCTIOllAL TESTS are not required when the systems are not required to bo OPERADLE or are tripped. It tests are missed, they shall be performed prior to returning the systems to an OPERABLE status.

(d) This instrumentation is exempted f rom the CllAll!1EL FUllCTIOllAL TEST definition (Dofinition 1.6). This CilAlillEL FU!iCTIO!!AL TEST will cloctrical signal into the consist of injecting a simulated measurement CllA!!!!ELS.

(e) Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to startup, if not performed within the previous 7 days.

(f) only the electronica portion of the thermal switches will be tested using an electronic calibrator during the throo month tost. A water column or equivalent will be used tc test the dp switches.

(g) With any control rod withdrawn. Not applicable to control rods removed per Specification 3.10.D or 3.10.E.

(h) This function is not required to bo OPERABLE when the reactor prossure vossol head is not bolted to the vossol.

-(i) Within one wook after entering OPERATIOllAL MODE 1 and then quarterly thorosfter.

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

3.1/4.1-11

l

. . j

+ ,  !

QUAD CITIES UllITS 1&2 DPR-29 & DPR-30 TABLE 4.1-1 (Continued)

BEACTOR PROTECTIOlt SYSTEld_(ECBAM) IllSTRUME!!TATIOll CllAliffEL FUllCTIOllAIuTEST ALID CHAllNEL Cl!ECK REOUIREMRIIf!

IABLE 110TATIOlis (k) The provisions of Specification 4.0.D are not applicable provided the CllAlll1EL FUliCTIOliAL TEST is performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after ontoring OPERATIOllAL MODE 2 from OPERATIOllAL MODE 1.

(1) Verify measured core flow to be greator than or equal to established coro flow at the existing pump spood.

l l

L I

3.J/4.1-12 l

o OUAD CITIES UNITS 1 & 2 DPR-29 & OpR-10 l

TABLE 4.1-2 j PEACT';R PPOTECTION SY1HM (SCRAM) I!!STRUME!!1AUM l CHAN!!EL CALI3 RATION REOUIPEMElll$

CHANNEL App 1&cao.w CAL 1BRATICN Mint.mem OPf.kAM CNAL

• rtp Tunctten Method (aI (f) freque.ey (b; WCES IBM High Flux Electronic R 2 1.

Calicratten R 3, 4, 5

2. APRM High Flux

a. Flow Bias Standard Pressure W(d)(k), R 1 and voltage Source

b. 15% Scram Electronic R 2 calteratten P 3, 5

c. Scram clamp Electrcnic W(d), R 1 Calleratton

3. LPP.M (h) Using T1P System (g) 1

4. React 3r High Standard Pressure R 1, 2(3)

Pressure Source

5. Drywell High Standard Pressure R 1, 2(1)

Pressure Source

6. Reactor Low Standard Pressure R(e) 1, 2 Water Level Source

7. Turbine Condenser Standard Vacuum source R 1 Low Vacuum

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

High Radiation Source (c)

9. Turbine INC Control Standard Pressure R 1 Fluid Low Pressure Source

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

11. HLgh Water Level in Standard Pressure A 1,2,5 Scram Discharge Source voluine (op only) 3.1/4.1-13 1

l l

1 l

t

• 4 l

QUAD CITIES UNITS 1&2 DPR-29 & DPR-30 TADLE 4.1-2 (Continued)

MOTOR PROT.ECIl0H_SLSTEM (SCRAM) IUHIJ1WiRITATl.0lfEQUJJiDiGILS CliAtli{EL CALIM1ATION REOUIREMENTE l

IAHLE NOTATIONS l l

(a) Neutron detectors may be excluded from the CllANNEL CALIBRATION. 3 (b) CHANNEL CALIBRATION tests are not required when the systems are not required to be OPERABLE cr are tripped. If tests are missed, they shall be performed prior to returning the systems to an OPERABLE status.

(c) A current source provides an instrument CHAN!!EL alignment every 3 months.

(d) This calibration shall concist of the adjustment of the APRM channel to conform to the power values calculated by a heat calance during OPERATIO!!AL MODE 1 when thermal power > 25% of RATED THERMAL POWER.

Adjust the APRM channel 11 the absolute difference is greater than 2% of RATED THERMAL POWER. Any APRM channel gain adjustment made in compliance with Specification 2.1.A or 2.1.D shall not be included in determining the shsolute difference.

(c) Trip units are calibra ed 6t least once por quarter and transmitters are calibrated at leart once per OPERATING CYCLE.

(f) Response time is rot part of the routine CHANNEL CHECK and CHANNEL CALIBRATION.

(g) Every 1000 equivalent full power hours.

(h) Does not provide scram-function.

This function is not required to be OPERABLE when PRIMARY (1)

CONTAINMENT INTEGRITY is not required.

(j) This function is not required to be OPERABLE when the reactor i pressure vessel head is not bolted to the vessel.

(k) This Calibration shall consist of the adjustment of the APRM flow biased channel to conform to a calibrated flow signal, 3.1/4.1-14

6 e QUAD CITIES UllITS 1& 2 DPR-29 & DPR-30 3,1 LIMITING CONDITIONS FOR OPERATION BASES The reactor protection system automatically initiates a reactor scram to:

a. preserve the integrity of the fuel cladding, I

b. preserve the integrity of the primary system, and

c. minimizo the onorgy which must be absorbed and provent criticality following a loss-of-coolant accident.

This specification provides the LIMITING CONDITIONS FOR 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 maintenanco. 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) . The system is made up of two independent trip systems, each having two subchannels of tripping devices. Each subchannel has an input from at least one instrument channel which monitors a critical paramotor.

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 causo a trip system trip. The outputs of the trip systems are arranged so that a trip on both systems is required to produce a reactor scram.

This system moots the requirements of the IEEE 279, " Standard for Nuclear Power Plant Protection Systems" issued September 13, 1966. The system has a rollability greater than that of a two-out-of-three system and somowhat loss than that of a one-out-of-two system (referenco APED 5179).

With the exception of the averago power range monitor (APRM) and intermediato range monitor (IRM) channels, each subchannel has at -loast one instrument channel. When the minimum condition for operation on the number of operable instrument channels por untripped protection trip system is mot. or if it cannot be mot and the affected protection trip system is placed in a tripped condition, the offectivonoss of the protection system is preacrved, i.e., the system can tolerato i a single failure and still perform its intended function of scramming the reactor. Three APRM instrument channels are provided for each protection trip system.

D 3.1/4.1-1

S t QUAD CITIES Ul1ITS 1 & 2 DPR-29 & DPR-30 APMs #1 and #3 operato contacts in one subchannel and APMs

1. 2 and #3 oporato contacts in the other subchannel. APMs #4,

1. 5 and #6 are arranged similarly in the other protection trip system. Each protection trip system has one more APM than is necessary to moet tho minimum number required por channol.

This allows the bypassing of one APM por protection trip system for maintenance, testing, or calibration. Additional IM channels have also boon provided to allow for bypassing of one such channel. The bases for the scram sottings for the IM, APM , high roactor pressure, reactor low water levol, turbino control valvo fast closure, and turbino stop valvo closure are discussed in Specifications 2.1 and 2.2.

Pressure sensing of the drywell is provided to detect a loss-of-coolant accident and initiate the omorgency core cooling equipment. The pressure-sensing instrumentation is a backup to the wator-lovol instrumentation which is discussed in Specification 2.1. A scram is provided at the same sotting as the omorgency coro cooling system (ECCS) initiation to minimizo the energy which must be accommodated during a loss-of-coolant accident and to provent the reactor from going critical following tho accident.

The control rod drive scram system is designed so that all of the water which is discharged from the reactor by a scram can be accommodated in the dischargo piping. A part of this system is an individual instrument volume for each of the south and north CRD accumulators. Those two volumes and their piping can hold in excess of 90 gallons of Wator and are the low point in the piping. tio credit was taken for those 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 dischargo 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 or no control rod insertion. To preclude this occurrence, level switches have boon installed in both volumes which vill alarm and screm the reactor when the volume remaining in either instrument volumo is approximately 40 gallons. For diversity of level sensing methods that will ensure and provido a scram, both differential pressure switches and thermal switenes have been incorporated into the design and logic of the system. -The setpoint for the scram signal has boon chosen on the basis of providing sufficient volume remaining to accommodate a scram, even with 5 gpm leakage per drive into the SDu. As indicated above, there is sufficient volume in the piping to accommodate the scram without impairment of the scram times or the amount of insertion of B 3.1/4.1-2

. -. ... . =- .- . _ - . . . - - - . . .- - -- - _

c> I QUAD CITIES UNITS 1 &2 DPR-29 & DPR-30 the control rods. This function shuts the reactor down while sufficient 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 proporly.

Loss of condonsor vacuum occurs when the condonser can no longer handle heat input. Loss of condensor vacuum initiates a closure of the turbino stop valves and turbine bypass valves, which climinatos the heat input to the condonsor.

Closure of the turbine stop and bypass valvos causos a proscure transient, neutron flux riso, and an increase in surface heat flux. To provent the cladding safoty limit from boing exceeded if this occurs, a reactor scram occurs on

-turbino stop valvo closuro. The turbino stop valvo closuro scram function alone is adequate to provent the cladding safety limit from being excooded in the event of a turbine trip transient with bypass closure.

The condenser low-vacuum scram-is a backup to the stop valvo closure scram and causes a scram before the stop valvos are '

closed, thus the resulting transient is less savoro. Scram occurs at 21-inchos Hg vacuum, stop valvo closure occurs at 20-inches Hg vacuum, and bypass closure at 7-inches Hg vacuum.

High radiation levels in the main steamline tunnel above that due to the normal nitrogon and oxygen radioactivity are an indication of leaking fuel. A scram is initiated whenever such radiation level exceeds fiftcon times normal 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. Dischargo of excessive amounts of _ radioactivity to the site environs is prevented by the air ejector off-gas monitors, which cause an isolation of the main condonsor of f-gas line provided the limit specified in Specification 3.8 is exceeded. 1 The main steamline isolation valvo closure scram is set to scram when the isolation valves are 10% elosed from full open.

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

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

Whenever the reactor modo switch is in the REFUEL or STARTUP HOT STANDBY position, the turbino condonsor low-vacuum _ scram and main steamlino isolation valve closure scrams are bypassed. This bypass has been provided for flexibility l

l B 3.1/4.1-3 l

a i QUAD CITIES UNITS 1&2 DPR-29 & DPR-30 during startup and to allow repairs to be made to the turbino condenser. While this bypass is in offect, protection is prcvided against pressure or flux increases by the high-pressure scram and APRM 15% scram, respectively, which are offective in STARTUP/ HOT STANDBY.

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

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 excessivo power levels and short reactor periods in the startup and intermediato power ranges (reference SAR Sections 7.4.4.2 and 7.4.4.3). During refueling, the primary Noutron Monitoring System (NMS)-indication of neutron flux levels is provided by the Source Range Monitors (SRM). The SRMs provido input to the RPS, but shorting links are installed across the normally closed contacts such that tripping an SRM channel does not affect the RPS. To activate the SRM scram function, those shorting links must be removed from the RPS. The SRM control rod scram providos backup protection to refueling interlocks and SHUTDOWN MARGIN should a prompt reactivity excursion occur. Although the IRM and APRM functions are required to bo OPERABLE during refueling, the SRMs provide the only on-scalo monitoring of neutron flux lovels during refueling and therefore the shorting links must be removed to-enable the scram function of the SRMs, The RPS (and thorofore 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 withdrawn from a core cell containing no fuel assemblies do

. not affect the reactivity of the core and therefore are not I

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 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 conditions, the capability of the one-rod-out interlock to prevent criticality has been demonstrated and the backup scram protection provided by the IRMs is sufficient to ensure a highly reliable scram if required. In the power range, the APRM system provides l B 3.1/4.1-4

. -- - = = _ -

e t QUAD CITIES UllITS 1 & 2 DPR-29 & DPR-30 required protection (roforence SAR Section 7.4.5.2). Thus, the IRM system is not required in the RUll OPERATIollAL MODE, the APP.Ms cover only the intermediato and power rangos and the IRMo provido adequato coverage in the startup and intermediato range. j The high reactor pressure, high drywoll pressure, low reactor water lovel scrams aro required for OPERATIollAL MODES 1 and 2.

The scram dischargo volume high lovel scram is required in OPERATIONAL MODES 1, 2 and 5. They are therefore required to be operational for thoso OPERATIO!1 AL MODES of reactor operation.

The turbino condonsor low-vacuum scram is required only during powor operation and must be bypassed to start up the unit.

B 3.1/4.1-5 I

. 4 QUAD CITIES UllITS 1&2 DPR-29 & DPR-30 4.1 SURVEILIA!1CE REQUIREME!JTS BASES A. Survoillance requireacnts for the reactor protection system are selected in order to demonstrato propor function and operability. The surveillanco intervals are datormined in many dif ferent ways, such as, 1) oporating experience, 2) good engineering judgemont, 3) rollability analysos, or 4) other analycos that are found acceptable to the liRC.

The frequency of calibration of the APRM flow-biasing network has been established at each refueling outage.

The flow-biasing network is functionally tested at 1 cast once por quartor and, in addition, cross calibration checks of the flow input to the flow-biasing network can be mado during the functional test by direct motor reading (IEEE 279 Standard for lluclear Power Plant Protection Systems, Section 4.9, September 13, 1966).

There are several instruments which must be calibrated, and will tako several days to perform the calibration of the entiro network. While tha calibration is being performed, a zero flow signal will be sont to half of the APRMs, resulting in a half scram and rod block condition.

Thus, if the calibrations woro performod during 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 significantt tho r o .'o r o , to avoid spurious scrams, a calibration frequency of each refueling outage is establiahod.

Reactor. low water level instruments 1(2)-263-U7A, 1(2)-263-57B, 1(2)-263-58A. and 1(2)-263-58B have boon 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 for the trip unit has boon established in General Electric topical Report 11EDC-30851P-A as quarterly. An adequate calibration / surveillanco test interval for the

-transmitter is once por operating cycle.

Specified surveillance intervals and surveil]ance and maintenance outage timos have been determined in accordance with 11EDC-30851P-A, "Tochnical Specification i Improvement Analysis for BWR Reactor Protection System,"

as. approved by the flRC in a letter dated July 15, 1987 from A. Thadani to T.A. Pickens 1

B 3.1/4.1-6 L

l l

e 4 QUAD CITIES UllITS 1& 2 DPR-29 & DPR-30 The turbino control valve fast acting solenoid valvo pressure switches directly measure the trip oil pressure that causes the turbino control valves to close in a rapid manner. The roactor scram notpoint was developed in accordance with llEDC-31336 "Gonoral Electric Instrument Sotpoint Methodology" dated Octobor,1986. As part of the calculation, a calibration period is inputted ,

to achiavo a nominal trip point and an allowable setpoint (Tochnical Specification value) . The nominal sotpoint is procedurally controlled. Based on the calculation input, the calibration period is defined to be overy Refueling Outage.

The sensitivity of LPRM detectors decreases with exposure to neutron flux at a slow and approximately constant rato. Changes in power distribution and electronic drif t also require compensation. This compensation is accomplished by calibrating the APRM system overy 7 days using heat balanco data and by calibrating individual LPRM's at least ovary 1000 equivalent full-power hours using TIP traverso data. 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 boon included in the latter table. Those are modo switch in shutdown, manual scram, main steamline isolation valvo closure, and turbino stop valvo closure. All of the devices or sensors associated with those scram functions are simple on-off switches, henco calibration is not applicable, i.e., the switch is oitner 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 dischargo volume tank. Based on the above, no calibration is required for those instrument channels.

B. The MFLPD shall be checked once per day to datormine 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 control rods are moved daily, thus the peaking factors are not expected -to change significantly and a daily check of the MFLPD is adequato.

l B 3.1/4.1-7

. i I 7

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

1. Licensing Topical Roport 11EDO-21617-A (Docomber 1978). .'

2. Gonoral Electric Topical Report fl0DC-30851P-A. _

3. IlEDC-31336 "Gonoral Electric Instrumont Sotpoint Methodology" dated October, 1986.

t

~

i r

B 3.1/4.1-8

..._.__-.,_.~,__.,_..,_.._.______.___..._.,,.,,_.._...a.._,_..__.~,...-,_ _, . _ . . . . _ .

EXISTING TECH SPEC l TS 3.1/4,1

' REACTOR PROTECTION SYSTEM' i F

f

e 4 l

QUAD C111:>

DPR 29 3.1/4.1 R[AC10R PR01[CT10N $Y51[H LIM 111NG COND1110NS FOR OPERA 110N $URV[IttANC[ REQUIR[M[Ni$

Applicability: Applicability:

Applies to instrumentation and associated Applies to the survelliance of the devices which initiate a reactor scram. instrumentation and associated devices which initiate reactor scram.

Objcctive: Objective:

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

SP[ClflCA110N5 A. The setpoints, minimum number of trip A. Instrumentation systems shall be systems, and minimum number of functionally tested and calibrated as instrument channels that must be indicated in Tables 4.1-1 and 4.1 2 operable for each position of the raspectively, reactor mode switch shall be as given in Tables 3.1 1 through 3.1 4. The systeu response times from the opening of the sensor contact up to and including the opening of the trip 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 when operating above 25% rated limiting power density (lif tPD) 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 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.

3.1/4.1 1 Amendnent No.114 t

I l

. 4 QUAD ClilE$

OPR 29

2. The power distribution shall be changed such that the maximum fraction of limiting power density no longer exceeds the fraction of rated power.

C. When it is determined that a channel is failed in the unsafe condition and Column 1 of Tables 3.1 2 through 3.1 3 cannot be met, that trip system must be put in the tripped condition immediately. All_other RPS channels that monitor the same variable shall be functionally tested within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The trip system with the failed channel .

may be untripped for a peripd of - **

time not to exceed I hour to conduct this testing. As long as the trip system with the failed 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 allow functional testing of all RPS instrument channels as specified 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 2 Amendment No. 114

1 o a QUAD CITl[$

OPR-29 3.1 LlHlilNG CONDlil0NS FOR OPERATION BASES The reactor protection system automatically initiates a reactor scram to:

a. preservetheintegrityofthefuelcladding 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.

This specification provides the Nd6'Q/gtMdWds'M/

/ MEr'aW// necessary to preserve the ability of the system to tolerate single f ailr;s and still perform its intended functior@ven during periods when instrurnent channels may be outofOervice because of maintenaYce. 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 7.7.1.29Thesystemismadeupoftwoidependenttripsystems,eachhavingtwo subchannels of tripping devices. Each suochannel 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 tw gic l.e., an input signal on either one or both of the subchinnels will cause a trip sys m trip. The outputs of the trip systems are arranged 50 that a trip on both systems is required to produce a reactor scram.

This system meets Abe requirements of the 1[EE 279htandard for Nuclear Power Plant Protection SystemP' issued September 13, 1966. The system 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 APID $179). +

t leas +)

With the exception of the average power rang monitor (APRM) and intermediate range monitor (IRM) channels, each subchannel has one instrument channel. When the minimum 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 system.

APRM(# 1 and # 3 operate contacts in gsubchannelandAPR[#2and#3 operate contacts in the other subchannel. APRFf f 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 minimum 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 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.

9 3.1/4.1-3 Amendment No. 114

~ _ - - - _ _ - - _ - _ _ _ . _ - - - _ - _ _ _ ____

. r

, QUAD *ClilLS DPR-29 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 1:,

discharged from the A partofthissystem/eactorbyascramcanbeaccommodatedinthedischargepiping.

1s an individal instrument volume for each of the south and north CR0 accumulatogn These two volumes and their piping can hold in excess of 90 gallons of water and.4btde 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 /eactor may not be accommodated which could result in slow scram times or partial of no control rod insertion. To preclude this occurrence, level switches have been installed in both volumes which will afarm and scram the /eactor when the volume remaining in either instrument volume is approximately 40 gallons. For diversity of level sensing methods that will ensure and 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 scragpven with 5 gpm leakage per drive into the SDV. As indicated above, there is suf ficient 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 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 valveg3yhich eliminates the heat input to the condenser. Closure of the turbine stop ar.d bypaYs valves causes a pressure transient, neutron flux rise, and an increase 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 alore is adequate to prevent the cladding safety limit from beint 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 G nches Hg vacuum, stop valve closure occurs at 2 O nches Hg vacuum, and bypass closure at O nches Hg vacuum.

3.1/4.1-4 Amendment No. lit

i

, s QUAD CIT][S OPR-29 High ediation levels in the main steamline tunnel above that due to the normal nitroges and oxyctn 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. ., scharge of excessive amounts of radioactivity to the_ site environs is prevented .,y the air ejector off gas t monitors, which cause an isolation of the main condenser off-gas line provided the limit specified in Specification 3.8 is exceeded.

The main steam 11ne 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.

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

WMnever the reactor mode switch is in the R N// or 5(//fdf/HM Sft/df 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 i repairs to be made to the turbine condenser. While this bypass is in effect, protectior is provided against pressure or flux increases b scram, respectively, which are effective in thiey the high mode pressureermooY.

eraerveAcr scram and APRM 15%

If the reactor N e M h t to a hot standby condition for repairs to the turbine condenser, the main steamline 1501stion valves would be closed. No hypothesized single failure or single operator action in this mode of operal. ion can result in an unreviewed radiological release.

The manual scram function is active in all modes, lcecearuwn.

thusmoces providing for a manual means of rapidly inserting control rods during all modes-of reactor operat4on.omar<evdL *o*M The IRM system provides protection against excessive power levels and short reactor periods in the startup and intermediate power ranges (reference SAP. Sections 7.4.4.2 and 7.4.4.3). - A-sovece-renge-monitoF(SRH)-system-is-elso-provided40-supply additional inser/ W -eevtcon--level-inforecticeducing-startup4vt4aeno-scram-funetten -ceference4Ag-Geet' en-Mr3rf)r---Thus-the-lR&ierequired4n-the M/il//-and-Sl/ H/f4f/p m/(k . In additionrprotection is-provided in this-range 4y-tha-AP -2W-scram)W-as-di scus sed-i n-t he4as e s-f ee4pec l44 c a t i on-P r h In the power range, the APRM system provides required protecti referencpSAR,Section7.4.5.2). Thus, the IRH system is not req Lred i in the // the APRM s~ cover only the intermediate and power ra

• the lR Fs provide adeq te c verage in the startup and intermediate range. ).

OPE t A rIONat.

higW8'rywell pressure,(rMwater level, and-scra The higNIeactor discharge-volume 4tph4 pressure,l-scrams-are-required-for-the4 eve tert / Hot-Standby of-plant-oper-attom lhey are therefore required to be operational for esejA///of reactor operation, enearecsat.

imMS AfC ftquiftk for cPEeQf TONAL intp[f5 l m l L , Th c gg m ry) doscharge volume hngh level bcrani is requirect in CPERar)cHAL rnen s 14.t. ma s.

1 3.1/4.1-5 Amendment No. 114 i

i,,____..._._ a._ __ - ._..--_ _ .-_u- -

_ _ . _ . . . _ . ~ - _ _ _ _ _ _ _ . . . . . _ . - - _ . . _ - . . - . - . . _ _ . _ . _ . _ .__._._~..__..m_._ _ _ _ . _. _. ._

e a r

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

The-requirement-that-the-1RH!$ -tie-inserted-in-the-core-when-the4PRHis-read 3/125-of full-scale-essures-that-there-is-properoverlap-in-the neutrvn-inonitoring Tystents and thus-that adeqJatetoverage-is provided-for-all-ranges-of -reacto& operation.

3.1/4.1-6 Amendment No. 114

G ..: c.

u QUAD-CITIES DPR-29 4.1 SURVEILLANCE REQUIREMENTS BASES A. e minimum functional testing frequency used in this specification it based a-. liability analysis using the concepts developed in Reference 1. This conc t was specifically adapted to _the one-out-of-two taken twice log c p. the reacto rotection system. Theanalysisshowsthatthesensorsarepry6rily ,

responsi e for the_ reliability of the reactor protection system.. Th s analysis a es use of " unsafe-failure" rate experience at conventio 1 t,nd nuclear powe plants in a reliability model for the system. An u afe failure

-is defined as e which negates channel operability and which, e to its nature,0 s revea d only when the channel is functionally tes d or attempts to respond to a reci gnal. Failures such as blown fuses, ru - ured bourdon tubes..faultedampl(tcrs,faultedcables,etc.,whichre o t in " upscale" or "downscale" readings o the reactor instrumentation are safe" and will be 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-respectingfunctional-testing.

These are:

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

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

-3. Device < which serve a us - ul function _only du ng sone restricted mode of_

operatun, such as St up/ Hot Standby, Refuel _ r Shutdown, or for which the only practical-t t is_one:that can be parfor d at shutdown (Group 3).

The sensors _that ma up Group-1 are specifically selecte rom among the whole iamily of industr on-off sensors'that have earned an~exc ent reputation for reliable op tion. Actual history on this class of sens s operating in

- nuclear powe ants shows four failures in 472 sensor years, o a. failure rate of 0.97 X 1 /hr. During design a goal of 0.99999 probability o success (at

• Lthe 50% co idence. level) was adcpted to assure that a balanced an dequate.

design _i achieved. :The probability of success-is primarily a funct of the M sensor allure rate and the test interval'. A 3-month test-interval was lanned -

for oup 1-sensors. This is in keeping with good operating practice an satisfies the design goal for the logic configuration utilized in the react r.

pfotectionsystem, r

Insert " 8 " J?-

_3.1/4.1-7 Amendment No. 114 1

, ~-

- - - --...-m-e * .-e-.-- 4- ,. ~#, , .-, - . . _ . . . , , . , , w--m -, - e

4 QUAD-CITIES OPR-29 TNsEtisfy the long-term objective of maintaining an adequate level of saf throughout the plant lifetime, a minimum goal of 0.9999 at the 95% cony de,ety nce level 4 proposed. With the one-out-of-two taken twice logic, thyl -(equires that each hosor have an availability of 0.993 at the 95% confidence level.

This level of h4, 1lability may be maintained by adjusting,tho' test interval as a function of the t(served failure history (ReferenceJ). To facilitate the implementation of thT5 technique, Figure 4.1-1 is,provided to indicate an appropriate . trend in test interval. The procedure is as follows:

1. Like sensors are poole None gpu the purpose of data acquisition.

2. The factor >l is the exposureMkand is equal to the number of sensors in a group, n,. times the elapsed time TgnT).

3. The accumulated nunpe f unsafe failureL is plottea as an ordinate against M as an absciss.a4n p

Figure 4.1-1.

4.

After a trendgoal satis,fy'fhe is established, will be thethe appropriatc test interval tomo%th)ly test the 1 f of theinterval to plotted pirits.

5. A test interval of 1 month will be used initially until a t d is established, c- ~

move m The turbine control valve fast acting solenoid valve pressure switches directly errcq l measure the trip oil pressure that cduses the turbine control valves to close in a LAsr rapid manner, The reactor scram setpoint was daveloped in accordance with NEDC Pa gocranpH g31336 " General Electric Instrument Setpoint l4ethodology" dated October,1986.As cal A n r part of the calculation, a calibration perfod is inputted to achieve a nominal trip FMs 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 tiery Refueling Outage. _

Grup2devicesutilizeananalogsensorfollowedbyanamplifierandabistple trip 'rcuit. The sensor and amplifier are active components, and a faipre is almost ahays accompanied by an alarm and an indication of the source, cT trouble.

In the-eventXfailure, repair or substitution can start immedija eTy, An as-is failure is one that " sticks" midscale and is not capable of ing either up or down in response to an ouMf-limits input. This type of fail for analog devices is a rare occurrence and is detectable by an operator who obs rves that one signal does not track the other three. Fo purposesofanalysjsfitisassamedthatthisrare failure will be detected withi( n t hours.

The bistable trip circuit which is he -cr the Group 2 devices can sustain unsafe l- failures which are revealed only on . 9herefnre, it is necessary to test them periodically. N A study was conducted of the nstrumentation channe1Qncluded in the Group 2 devices to calculate the V ' unsafe' failure rates. TheT nalog devices (sensoy and amplifiers) are prejkfed to have an unsafe failure rate ofMe(s than 20 X 10 failures / hour. T#ebistQletripcircuitsarepredictedtohave unsafe failure rate of lestAh,an 2 X 10 failures / hours. Considering the 2-hour itoring

! intervaj for the analog devices as assumed above and a weekly test int 1 for the bistabTe trip circuits, the design reliability goal of 0.99999 is attaine (d with amplem9rgin.

3.1/4.1-8 Amendment No.129

e ..-

t y0AD ClilES OPR 29 T he-bi s t abl e-< lev ic et-a m mon i to red- duri ng - pl ant- ope ra ti on-to-record -thei r- f a i l ure

- history-and-establish-e-test-4ntervel--using-the-curve-ofiigure 4rl-IrThere are numerous-4 dent 4 cal-bistablo-devices-used-throughout-the plant-instrumentation systemrThereforer-signifIcant-data-on-the-f ailure-rates for theAlstable_detices

- shouldle_accumulat edlapidly.

(gaarter)

The frequency f calibratlen of-the APRM flo G lasing network has been established

-at each refu ing outage. The flo6 biasing network is functionally tested at least once per mon 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 U tire network. While the cali tion is being performed, a zero flow signal will be sent to half of the APR esulting:in a half' scram and rod block condition. Thus,ifthecalibratio@

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

- floGiasing networigis not- significant; therefore, to avoid , spurious scrams, a calibration frequency of each refueling outage is established.

(M - ca ca (n-Reactor low water-leve1' instruments 1-263-57A', 1-263-578, 1-263 58A, and 1-263-58B 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 'runet40nel-test 4ng-ferr instrument-4 oops-of-the-analog-trip-systear -including-reactor !w water __1cuelJas.

[@ been estabtished in44cen&4ng-Topical-Repor-tJE00=2-16172A-(DecembeF-19 one=outiofitwetaken-twice-logie -MEDO-2161LA-stetet-that-eachr-trip-unit-be r

' subj ec t ed 4 e-a-t+14 bre t4enh unc tion a l-t es t-t>f-ene- mon t h . An adequate calibration / surveillance test interval for tne transmitter is once per operating _

cycle.

I, J.4 pee.ifieel wrveillance inicevals and sarveillance ancL maihlena&1cc=

oudage +imes have been ciclermined. in accorcia nce. wilh NEDC ~ 3CB sit'- A s

'

• Technica.1.Specif:calien Irnprovemenf Anahjses for au)R Reaclor Pre lec t en Sjsecs"(as agreveel by the NRC in a leHer

~*~

^

cia.tect JuJy ir,199 9 from A; Thada ni to r A, Piclecr\s .

[ Add paragraph irorn= page -S I/ 4.I-e ]

for fhe ' trip unit ha s b'cen eslablishcl in fenerctl El c}r'ic.

Topica.I Reporf ^tEDC- MB51P-A as_ tjuarf erly, o

L I

3.1/4.1-9 Amendment No 114 l-

a m QUAO-CITIES OPR-29 Grqup 3 devices are active only during a given portion of the operatiort . For examp Thus,(thethe ly IRM test thatis active is meaningful during startup andperformed is the one inactive during j St prior fu'l196Cer operation.

to shutdown or startup, , the tests that are performed just prior se of the instrument.

Calibration frequency the instrument channel is ded into two groups. These are as follows: ,

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

2. Vacuum tube or semicon . or devices and de stors that drift or lose sensitivity.

Experience with e sive type instruments in Commonwealth Ediso nerating stations and substat indicate that the specified calibrations are adeq (uais For those devices h employ amplifiers, etc. drif t specifications call for dhtto be lest tha3A / month 1.e. , in the period of a month a drif t of 0.4% would occur, uut

. providing for adequate margin.

The sensitivity of LPRM detectors decreases with exposure to neutron flux at a slow and approximately constant rate. Changes in a power distribution anJ electronic drif t also require compensation. This compensation is accomplished by calibrating the APRH system every 7 days using heat balance data by calibrating individual LPRM's at least every 1000 equivalent full power hours usir.g TIP traverse data.

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 steamlino isolation valve closure, and turbine stop valve closure. All of the devices or sensors associatedj with these scram functions are simple ov off switches, hence calibration is not applicable, i.e. , the switch is either on or of f. Further, these switches are mounted solidly to the device and hi.vc 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 MFLPD shall be checked once per day to determine if the APRM scram requires adjustment. TLis may normally be done by checking the LPRM readings, TlP traces, of process computer calculations. Only a small number of control rods are moved daily, thus the peaking factors are not expected to change significantly and a daily check .

of the MFLPD is adequate.

References

-1. I r &-J acobs7-MeH aM M ty-o f-E ngi nee re44a f ety4 ea ture s-a s-a -F unc t i on-o f Testief,4requeney"7-Huelear-Safety;-Vol-trNo-47pF310 317;~3UlyvAligust 1%8, I 'A Licensing Topical Report NE00-21617-A (December 1978).

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

2. General Eleciric. Tcpical Referi NEOc.-3C85IP ~ A 3.1/4.1-10 Amendment No.129

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l INSERT FOR TECHNICAL SPECIFICATION SECTION 3.1/4.1 REACTOR PROTECTION SYSTEM" Insert "A" 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 closed contacts such that tripping an SRM channel does not affect the RPS. To activate the SRM scram Iunction, 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 during refueling, the SRMs provide the only on-scale monitoring of neutron flux Levels dunng 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 fueiassemblies. Control rods withdrawn from a core cell containing no fuel assemblics 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 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 conditions, the capability of the one-rod-out interlock to prevent criticality has been demonstrated and the backup scram protection provided by the IRMs is sufficient to ensure a highly reliable scram if required.

Insert "B" .

L L 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 ) other experience,2) good engineering udgement,3) reliability analyses, or 4 analyses that are found acceptable to the NRC.

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SIGNIFICANT HAZARDS CONSIDERATIONS

' AND ENVIRONMENTAL ASSESSMENT EVALUATION

-PROPOSED TS 3.1/4.1 REACTOR PROTECTION SYSTEM" 4 aDr- -+ ^ W'-+F se wa-%d =>,>-w -w m 4' Du + r N rT'e m- gwsyog- w- w- = w w ei == y y 9--e,.am'. T--y e s- y- y+ -yv -mm7-pt+y"tmst-'-y'e- e v v V4' * - -

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EVALUATLQH FSB BIGNIFISANT HA1ARDS. CONSIDERATION PROPOSLD SPECIFICATION 3.1/4.1 REACTOR PROTECTION SYSTEM 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 reviewed by Commonwealth Edison and we believe that they do not present a Significant Hazards Consideration. The oasis for our determination is documented as follows:

BASJE f.QB ILQ SIGNIFICALLT HAZARDS CONSIDERATION Commonwealth Edison has evaluated this proposed amendment and determined that it involves no significant hazards 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:

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,

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 ar~

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 pirement 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 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.

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

o* 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,

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 Determination of No Significant Hazards for the proposed change. Implementation of the extended surveillance Calibrationsthe intervals for willChannel Functional not be made withoutTests 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, 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 respect to the APRM/IRM companion trip are the Rod Drop 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 evaluated accidents.

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

a. Since the Generic Changes proposed to the technical specifications are administrative in nature, they cannot 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 requirements with existing Quad Cities requirements to provide a user friendly format and 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 to the Tables follow later operating BWR guidelines that are presently being utilized at these plants and have been

v 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,

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 by these changes. No new modes of plant 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 oxtensions.

d. The deletion of the APRM Dow:'. scale Scram Trip Function does not introduce any new accident scenario. The 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 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 the initial power level could be a few percent lower which would not have a significant effect on the severity of the event. In addition, proper overlap between the IRMs and APRMs is not affected since the calibration requirements are not being changed.

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

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 3.1.7/4.1.A and 3.1.B/4.1.B implement an STS type of format while retaining the present twc .olumn 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 been evaluated and found to be acceptable for use at Quad Cities. The deletion of present Surveillance Requirement 4.1.C does not involve a l

l-

, . , _ _ _ _. . ._ _ . . . _ _ _ - . . ~ . . - -._ -._ _ _ . _ .. . . .

N significant reduction in the margin of safety since other equivalent methods are utilized to determine if the

- failure 1of one.RPS channel in an unsafe position affects o',herJsimilar channels.

The changes to the Tables in Section 3.1/4.1 follow proven 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 cother changes to the reduction tables involve =in the margin of safety,inor clarifications or m improvements that do not affect the margin of safety.

c. The 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 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 foundracceptable 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 setpoint calculations before implementing the-extended surveillance testing intervals for both -the ' Channel Functional Tests and the Channel Calibration Tests.

Therefore, the changes do not involve aJsignificant-Ereduction in the-margin of safety.

- d. The APRM Downscalo Scram Trip Function-is not credited in the termination of any-FSAR or reload-safety analysis event. Astsuch,-the elimination-of this scram function

- has noleffect on any margin of safety.

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w ,x ENVIRONMENTAL ASSESSMENT EVALUATION PROPOSED SPECIF7. CATION 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.

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 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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QC-1/ QC-2 DIFFERENCES TS 3.1/4.1

' REACTOR PROTECTION SYSTEM'

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

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

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 Technical Specification-Bases section 3.2 contains the background for this statement.

Several administrative differences were identified as follows:

Pace 3.1/4.1-1 Applicability Unit 1: Applies to instrumentation and ...

Unit 2: Applies to the instrumentation and...

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

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

Page 3.1/4.1-4 Paragraph 2 Unit 1: into the SDV. As indicated above, there...

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

Paragraph 3 Unit 1: Loss of condenser vacuum...

Unit 2: Loss of condensate vacuum...

sy n face 3cl/4.1-5 Paragraph 7 Unit 1: discharge velume 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 ...

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

Unit 2: rare occurrence and is detectable by an operator who observes than on signal....

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

Unit 2: amplifiers) are predicated to have..

Paragraph 4 Unit 1: failures / hour. The bistable trip circuits are predicted to have ...

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

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 attained with ...

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

Paraaraoh 3.1/4.1-9 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 ...

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

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

1

w effL Paqe 3.1/4.1-10 Paragraph 1- Unit 1: Group 3 devices are active only during a given portion of the operation cycle ...

Unit 2: Group 3 devices are active only uuring a given portion of the operational cycle ...

Paragraph 3 Unit 1: and substations indicate that Unit 2: and substations indicates that Paragraph 4 Unit 1: and approximately constant rate.

Changes in a power distribution ...

Unit 2: and approximately constant rate.

Changes in power distribution ...

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

Unit 2: the APRM system overy 7 days using heat balance data and by calibrating Pace 3,1/4.1-15 Note (9) Unit 1: electronic calibrator during the three month test ...

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

t-