ML20154K910

Proposed Tech Specs,Clarifying Rod Block Monitor Operability & Bypass Time Requirements

ML20154K910
Person / Time
Site:	Quad Cities
Issue date:	05/17/1988
From:	COMMONWEALTH EDISON CO.
To:
Shared Package
ML20154K906	List: ML20154K903 ML20154K910
References
NUDOCS 8805310085
Download: ML20154K910 (20)

Text

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A i *s 3.2/4.2 *RO?EC?!VE IN5faVMEN?A?!CN LIM!?!NG CCN0!TIONS FCR OPERAf!ON suave!LLANCE aEQUIREMEN?$

.' 'Agdlicability:

• ApollCability:

Applies to the plant Instrumentation Applies to the sarveillance requirements which performs a protect 1v'e funct* ion. of the instrunwntation that performs a protective function.

Objective:

Objecttve:

To assure the operability of protective To specify the type and frecuency of sur-instrumentation. veillance to be apDlled te protective Instrumentation.

A. Primary Contalnaent Isolatten Func.$PECIFICATIONS tions A. Primary Containment Isolation Func-tions When primary contairtent integrity is required. the limiting conditions of operation for the Instrumentation Instrumentation and logic systems that initiates artmary contatement shall be functionally tested and cal-1 solation are given in Table 3.2-1. ibrated as Indicated In Table 4.2-1 B. Core and Containeent C oling systems - B.

Initiation and Control Core and Containment Cooling Systems -

Initiation and Control The limiting c0nditions for Oceration

'or the Instrumentation that Instrumentation and logic systems nitiates or centrols the : Ore and shall be functionally tested ind containment cooling systems are given calibrated 4.2-1. as indicated in Taole In Taole 3.2-2. This Instrumentation must be ocerable .nen the system (s)

It Initiates or centrols are ee:uireo to be ccerable as specified in Spect-ficaticn 3.5.

C.

Control Rod 91cck Actuation C.

Centrol Rod Block Actuation 1

The limiting conditions of ocer-ation for the instrumentation Instrumentation and logic systems that initiates centrol re: clock shall be functionally tested anc cal-are given in Table 3.2-3. ibrated as indicated in Table 4.2-;

2. a. When a Limiting Control Ecd Pattern exists one of the Red Bleck Monitors may te bypassed for maintenance and/or testing previce: that this cencit ten does 9:t last lenger than 24 neurs in a 30 day period. If tnis condition lasts 10nger thar 24 nours In a 30 cay :er10d the systen shall Oc tri ced.

The L tTe s:ent .ntle in a timiting C:ntrol 3 d 8attern with cre r more 4:0 31 ct Montters cycassec r ine:eracle anc :c withcrawal 21:cxt: 30es 9et cunt against ne 24 grs in a 30 Oay per100.

l

b. One channel may te Sy:assed acove 30*, : er without a t ce restriction :roviced that a L1miting C:ntr:1 20d 8attern Oces not esist and the remaining 400 310cx Monitor channel 15 c era:le.

! c. Both R0d Block wenitor Channels are autceatically bypassed at less than 30*

rated thermal pcwer or tf the selected centrol r:: nas one or more adjacent fLe1 l bundles cceprising the cuter boundary of the reacter : ore.

Il39H 3.2/4.2-1 l amenccent No.

8805310085 880517 PDR ADOCK 05000254 i P DCD

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

A QUAD-Cl?!ES i- ,* OpR 20 steam fion1tneanG break also accident.Itmittag the loss of mass inventory from the ve

• ng

• Guring a steam Instruawntation is provided .ntch causes a trip of Group 11In valves. addition to monitoring steam flow.

solation the main steamline. thus only GroupThe primary function of the instrumentation is to detect a 1 valves are closed.

accicent, 140% of ratedmain steamitne steam flow, break outside the drywell, this trip setting ofFor the orst-case tn conjunction with the flow limiters and euin steamline valve closure, limits the mass inventory loss such that fuel is not uncovered. fuel temperatures remain less than 1500 0 F, and release of radioactivity to the environs is well below 10 CFR 100 guidelines (re?arence SAR Sections 14.2.3.9 and 14.2.3.101.

tunnel to detect lears in this area. Temperature monitoring instrumentation is provided in Trips are provided on this instrumentation valves. andof It: setting when2000 exceeded F cause closure of Group I tsolation of 5 to 10 gam: 15 low enough to detect leaks of the order breaks. For large breaks, it 15 a backup to high-steam flowthus it ts cacable of coverin instrumental on discussed above, and for small breaks with the resulting small release of radioactivity, exceeded. gives isolation Oefore the gutcelines of 10 CFR 100 are detect gross fuel failu e.High radiation monitors in the main steamline tunnel have been provided to valves, the only valvet This instrumentation causes closure of Group 1 required to close for this accident. Wtth the estabitshed setting of 7 times normal background and main steamline 100 guidelines are not exceeced for this accicent (reference SAR Sect 12.2.1.7).

Pressure instrumentation is provided nich trips when main steamitne closure ofcrocs pressure Groupbelow 325 estg.

1 isolation valves.A trip of this instrumentation results in modes this trip function 15 bypassed. In the Refuel and Startup/Het Standby to provide protection against a cressure regulator malfunction antch wouldThis function is cause the control and/or bypass valve to open. With the trip set at 325 psig. Inventory loss is limtted 50 that fuel not uncovered and peak cladding temperatures are much less than 1500 0is F. thus, there are no fisston products available for release other than those in the reactor water (reference SAR Section 11.2.3).

The RC1C and the HPCI high flew to detect a Dreak tn their respective and temperature Instrumentation are provided piping. Tripping of tMts instrumentation valves. results in actuation of the GC1C or cf NPCI 1solat10n Tripoing logic for this function is the same as that for the main steamline isolation valves, thus all sensors are respired to ce 0;eraole or in200*F of a trip;4d ccndition to meet single-failure crite-ta.

and 300% of cesign flow and valve closure time are sucn that coreThe trip settings uncovery is o*evented and fisston product release is within limits.

i The instrumentation .nien intttates ECCS action is arranged in a one-out-of-t.o taken twice icgic circutt, circuits, ho-ever, Unlike the reactor scram there is one trip system associated with each function rather than the t.o trip systems in the reacter protection system. The single-failure cooling criteria functions are met by virtue of the fact that re:Undant core are provided, e.g. , sprays and automatic blewdc.n and 1 nigh-cressure system tec:mes coolantincceracle, injection.

the systemThe.nich specification requires that if a trt:

inoceracle. : r evarele, if tne trip system for it activates ts declared core spray a ceccmes ineceracle, ccre specificattens of spray a is :eclared inoceracle and the Out-of-service the effectiveness of Scecthei system f i cat s et.

.ith3.5 ;overn. This s:ecificatten creserves res:ect to the single-failure criteria even during :ertecs anen matntenance or testing is cetng perfor ed.

j 1139H L 2/4. 2 -6 A : e n dt.;e n t NO.

a

l i ,

a CUAO-CITIES 0@R-29

. , .' . Thd contrdl 706 bloch functions are provideG to prevent excessive control

. rod withdrawal so that MCPR does not go below the MCPR Fuel Cladding Integrity Safety limit. The trip log 1C for this function Is one out of n; e.g., any trip on one cf the 514 APRM's. eight IRM's, four SRM's will result in a rod block. The minimum Instrument channel reouirements assure suffiClent instrumentation to assure that the single-fallure criteria are met. The mint:rtam Instrument channel requirements for the RBM may be reduced by one for a short period of time while in a Limiting Control Rod Pattern to allow for maintenance, testing, or calibration. This time period is only 3*. of the operating time in a month and does not $1gnif1Cantly increase the risk of preventing an inadvertent control rod withdrawal. In addition.

While the unit is operating in a limiting control pattern with one or more Rod Block Monitors bypassed and control rod withdrawl blocked. this time does not count towards the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> In a 30 day restriction. This time restriction 15 placed on the Rod Block monitor system to decrease the probab111ty of a Rod Withdrawl Error while in a limiting control rod pattern. With control rod withdrawl blocked all rod withdrawl 15 prevented, hence the RBM is not required to function.

l ll39H 3.2/4.2-64 Amend:nen t NO.

g' o QUAO-CIf!!S i .* OPR-29 The'APRM rod block function is flow biased and prevents a signiftcant reduction In MCPR especially durtng operation at reduced flow. The APRM

,srovides gross core protection, t.e.

rods in the normal wtthdrawal sequenc,e. limits the gross withdrawal of control In the set refuel at 12% of rated and power.

startup/ hot standby modes, the APRM rod block function is-protection in the Refuel Thts control rod block provides tne same type of and Startup/ Hot Standby modes as the APRM flow-Otased rod clock does in the Run mode, withdrawal before a scram ts reached.

t.e., prevents control rod The RBH rod block function provides local protection of the core, i.e ., the rod withdrawal error from a ilmiting control rodThe flow biased.

pattern. prevention of transition bo trto point is for each reload te assure that, with the specific trip settings, rodThe worst-ca withdrawal safety 11mit.is blocked before the MCPR reaches the fuel cladding integrity rod block action will not violate the fuel cladding Integrity safety 11m Thus the RBM rod block function 15 not requtred below this power level If .

5 a contrc1 red is selected that has one or more adjacent fuel bundles .

comprising the outer boundary of the reactor core, the neutron leakage is sufficiently cladding htgn such integrity thatlimit.

Safety withdrawal of this rod will not violate the fuel control outer rods that boundaries have of the core.one or more adjacent fuel bundles comprising theThus the RB The IRM block function provides local as well as gross core protection. Th e scaling 10 abovearrangement the Indicatedislevel. such that the trip setting ts less than a factor of Analysis of the aorst-case accident results in rod Itmit.

safety block action tefore "CPR approaches the MCPR fuel Cladding integrity A isdownscale or not senstttveindication enougn. on an APRM ts an indication the instrument has failed to changes in control rod motion, and the centrol rod motton is thusIn etther case tne ins prevented.

The downscale trips are set at 3/125 of full scale.

Vhe SRM rod block with i 100 CPS and the detector o un t f ll that the IRM's withdrawal for are startup.not withdrawn from the core orter to conmenctng rodinserted assures provide annunciation for operator action.The scram discharge volurre hign water level block The alarm setpoint has been selected level to provide adecuate time to allow determination of the cause of increase and corrective action prior to automatte scram initiation.

For ef fective emergency core cooling for small pipe breaks the HPCI system <

must allow function either core since reactor spray pressure or LPCI does not to operate decrease rapidly enougn to in time.

relief function is provided as a backup to the MPCI in-the event the HPCIThe autcmatic pressure does not operate. The arrangement of the tripping contacts is such as to provide this function . hen necessary and minimize spurious operation. The trto settings given tn the scecification are adecuate to assure the above criterta are met (reference SAR Section 6.2.6.3). The spectf tcation preserves the effectiveness of the system during periods of matntenance, testing or calteration and also mtntm12es the risk of inadvertent coeratten t.e.,

only one instrument enannel out of service. ,

Two radiaticn mentters are provided on the refueling ficor wnich initiate Isolation ofThe systems. thetric reactorlogicOutiding andofoperation ts one out two. Trip Of the standby gas treatment settings of 100 mR/hr for the monitors on the refueling ficor are based uoen inittattng normal ventilation isolation an: standby gas treatreent system operation 1139H 3.2/a.2-7

• t 1

I OPR-29 ]

4' ,

. TABLE 3.b 3 INSfaVMENTA?!CN ?HA? INITIATES 200 BLOCK

, , # Min 1murff NumbeG" .

of Operable or Tripped Instrument Channels per Trio system L1 Instru m t Trio Level settina 2 APRM upscale (flow blas){7) 1(0.58Wo + 50] MP (2]

MFLPD 3

2 APtM voscale (Refuel and ,112/125 full scale

$tartup/ Hot Standby mode) 2 APRM do.nscaleI7} 13/125 full scale 1

Rodclockynitorupscale(flow 014s 3 10.65Wo + 43N 1

Rod clock monitor downs: ale [7] 13/125 full scale 3 IRM downscaleIII {83 13/125 full scale 3  !RM upscaleI83 1108/125 full scale 2[5] t in Startup 12 feet below core centerline SRM detector {

posttton 3

IRM detector8;(ngt in Startup position 22 feet below core centerline 2(5) (6] SRM upscale tiOS coyng3y3,e 2[5] SRM downscale (9} 2102 counts /sec 1 (per bank) Htgh water level in scram i 25 gallons (per bank) discharge volume (50V) 1 50V hign water level scram NA teto typassed Notes

1. For the $tartup/ Hot $tanoby and Run Oositions of the reactor mode selector swittn, there shall be two operable or tr1cced trip systems for each function except the SRM rod blocks.

IRM uoscale and 129 : *nsca'e need not ce ocerable in the Run position, aPRM downsCale.

APRM upscale (f!ow biased). and RBM downscale need not be operable ir. the Startup/ Hot Standby N de. The RBM ucscale and downscale trips need not te c:erable at less than 30%

rated t9ermal power, or at any :ower level if the selected control red nat. one or more '

adjacent fuel bundles compr151ng the outer boundary of the reactor core. The RBM is automatically eypas.*,u at less than 30% rated thermal power or if the selected c:ntrol rod has One or tore ad; Cent fuel bundles c0frorising the outer boundary of the ructor core.

For systems with more than one channel ;er trip system. If the first column cannot te met for one of the two trio systems, this condition may exist for up to ? days =roviced that curing that time the coerable system is functionally tested it'Taediately and daily thereafter: if this c:ndition lasts longer than 7 days the system shall te triogec. If the first column cannot te met for both tric systems. the systems sna11 te tripped.

2. Wo is tne :ercent of cr1ve flow recutred to produce a rated core flow of 90 91111cn Ib/hr.

Trip level setting 1s in ;ercent of rated Ocwer (2111 Wt).

3. IRM :ownscale may te typassed .nen it is on its lowest range.

4 This funct10n is cycassed .nen the c:unt rate is GT/E 100 CPS.

5. One of the f:ur SRM tecuts may be cypassed.

6. fhis SRM functioft may te bytesseo in the M1gner IRM ranges tranges 5,9 ano 101 anen the !EN upscale rod clock is c:erable.

7. Not recuired to be oceracle .hile cerforming low power physics tests at atmoscneric pressure during or after refueling at power levels not to exceed 5 Nt.

3. This IRM function occurs when tne reactor mode switch is in the Refuel er Starty / Hot Standby position.

9. This trig 15 bypassed nen the SRM is fully Inserted.

1139H 3.2/4.2 14 Amendment NO.

e . QUAO- ITIES UPR-29

, '3 he control rod Grive housing 1. The correctness of the control

. support systen shall be in place rod withdrawal sequence input to during reactor power operatton tne RWH computer shall be vert-and wnen the reactor coolant fled after loading the secuence.

System 1$ pressurized above atmosoneric pressure with fuel in the reactor vessel, unless Prior to the start of control red withdrawal towards critical-all control rods are fully Ity, the capa0111ty of the rod inserted and Spectf1 cation worth minim 12er to properly ful-3.3.A.1 1$ met. fill its function shall be vert-fled by the following checks:

a. Control rod withdrawal se-quences shall be estae- a. The RWH computer on1 tne 11shed so that maximum diagnostte test shall ce reactivity that could De successfully performed, added by dropout of any in-crecent of any one control b. Proper annunctation of the blade would be such that the selection error of one out-rod drop accident design of-sequence control rod limit of 280 cal /gm 15 not shall be verified, exceeded.

5. Whenever the reactor is in c. The rod block function of the Startup/ Hot Standby or the RWM shall be ver1f ted my Run mode below 20% rated withdrawing the first rod as thermal power, the rod worth an out-of-secuence control minimizer shall ce rod no more than to the operacle. A second op- block point.

erator or qua11fted tech-nical person may be used as a. Prior to control rod withdrawal a substitute for an incoer- for startup or during refueling, aole red worth minimi:er verify that at least two source antch fails after with- range channels nave an observed drawal of at least 12 con- count rate of at least three trol rods to the fully counts per second.

withdrawn position. The r0d worth mintmt:er may also Oe $. When a 11miting control rod pat.

Dypassed for low cower tern exists, an Instrueent func.

pnysics testing to tional test of the RBM shall be demonstrate the snutdown performo.c ;rior to withdrawal of margin reoutrements of the des gnated rod (s) and datly Scecification 3.3.A If a theraafter.

nuclear engineer is present and verifies the step- 6. The scram discharge volume vent by-step rod movements of the and drain valves sna11 de vert-test procedure. fled open at least once :er 31 days. These valves may be 4 Control rods sna11 not be witn- closed tntermittently for test-drawn for startup or refueling Ing under acministrative c:ntrol unless at least two source range and at least once per 92 days.

cnannels nave an Ocserved ccunt each valve shall ce cycled rate equal to or greater than through at least one c molete three counts er second and cycle of full travel. At least these SRM's are fully inserted. ence each Refueling Outage the scram discharge < 1ume vent and S. Except as provided ey drain valves 4t11 te Scecification 3.2.C.I and demonstrated to:

3.2.C.2 during c:eration with limiting centrol r:d Oatterns, a. Close attnin 30 se: ncs af-as :etermined 3y the nuclear ter receipt of a signal for engtreer either: control red; to scram, and

a. b:th REM cnanrels shall te b. Ocen anen the scram s1gnal c;eracle. ts reset.

c. c:ntrol red withdrawal sna11 Oe 01ccred; or A nendment No.

1139H 3.3/a.3-3

g QUAD-CITIES i

DPR-29

. 'b.. t,he dJlayed neutron fraction chosen for the bounding reactivity curve

c. a beginning-of-life Doppler reactivity feedback

. d. scram times slower than the Technical Specification rod scram insertion rate (Section 3.3.c.1)

e. the maximum possible rod drop velocity of 3.11 fps

f. the design accident and scram reactivity shape function, and

g. the moderator tenperature at which criticality occurs In most cases the worth of insequence rods or rod segments in conjunction with the &ctual values of the other important accident analysis parameters described above, would most likely result in a peak fuel enthalpy substantially less than 280 cal /g design limit.

Shoulc a control drop accident result in a peak fuel energy content of 280 cal /g. fewer than 660(7 x 7) fuel rods are conservatively estimated to perforate. This would result in an offsite dose well below the guideline value of 10 CFR 100. For 8 x 8 fuel, fewer than 850 rods are conservatively estimated to perforate, with nearly the same consequences as for the 7 x 7 fuel case because of the rod power differences.

The rod worth minim 12er provides automatic supervision to assure that out of sequence control rods will not be withdrawn or inserted; i.e. it limits operator deviations from planned withdrawal sequences (reference $AR Section 7.9). It serves as a backup to procedural control of control rod worth. In the event that the rod worth minimizer is out of service when required. a licensed operator or other qualified technical employee can manually fulfill the control rod pattern conformance function of the rod worth minimizer. In tnis case, the normal procedural controls are backed up by independent procedural controls to assure conformance.

4. The source range monitor (SRM) system performs no automatic safety system function. i.e., it has no scram function. It does provide the operator witn a visual indication of neutron level. This is needed for knowledgeable and ef ficient reactor startup at low neutron levels. The consequences of reactivity accidents are functions of the initial neutrcq flux. The requirenent of at least 3 counts per second assures that any transignt. should it occur.

begins at or above the initial value of 10-" of rated power used in the the analyses of transients from cold conditions. One operable SRM channel would be adequate to monitor the approach to criticality using homogeneous patterns of scattered control rod withdrawal. A minimum of two operable SRM's is provided as an added conservatism. -

5. The Rod Block Monitor (RBM) is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high power operations. turing reactor operations with certsin limiting control rod patterns. the worst-case withdrawal of a single control rod could result in one or more fuel rods with MCPR's less than the MCPR fuel cladding integrity safety limit. During a Limiting control rod pattern, testing of the RBM system will assure its operability prior to withdrawal of such control rods. To facilitate testing while in a liaiting control rod pattern one RBM may be bypassed, for brief periods of time to perfor.n maintenance and/or testing without decreasing the reliability of the system, provided the other RBM is operable. Two RBM chanr.els are provided. Tripping one operable channel will block erroneous rod withdrawal soon enough to prevent violation of the MCPR Safety limit. It is the responsibility of the nuclear engineer to identify these limiting control rod patterns and the designated rods either when the patterns are initially established or as they develop

• due to the occurrence of inoperable control rods in other than limiting pattern.

6. The operability of the Scram Discharge Volume vent and drain valves assures the proper venting ar.d draining of the Volume, so that water accunulation in the Volume does not occur. These specifications provide for the periodic verification that the valves are open, and for the testing of these valves under reactor scram ccaditions during each Refueling Outage.

1139H 3.3/4.3-9

r w a ,o O ,

LIMI?!NG CON 0!?!ONS FOR OP(2A?!0N3.3/4.2 CRO?EC7!V2 INSTRUME4?a?!0N SURVE!LLaNCE ME0VIREMEN75 Appl t'ca b i l i ty: '

applicability:

Applies to the plant Instrumentation Applies to the surveillance requirements which performs a protective function, of the instrumentation that performs a protective function.

Objective:

Objective:

To assure the operability of protective To specify the type and frequency of sur-instrumentation. vet 11ance to be applied to protective tnstrumentation.

A. Primary Containment Isolation func-SPECIFICAT!ONS tions A. Primary Containevnt Isolation Func-tions When primary containment integrity is recutred, the limiting conditions of operation for the instrumentation Instrumentation and legte systems that initiates primary contatnment shall be functionally tested and cal-isolation are given in Table 3.2-1. ibrated as indicated in Table 4.2-1.

8. Core and Containaent Cooling Systems - 8.

Initiation and Control Core and Containaent Cooling Systems -

Initiation and Control The limiting conditions for operation for the instrumentation that Instrumentation and logic systems initiates or controls the core and shall be functionally tested and containnent cooling systems are given calibrated 4.2-1. as indicated in Table In Table 3.2-2. This instrumentation must be operable when the system (s) tt initiates or controls are recuired to be operable as specified in Spect-f icat ion 3. 5.

C. Control Rod Block Actuation C.

Control Rod Block Actuatiosi

1. The limiting conditions of oper.

ation for the instrumentation Instrumentation and logic systems that initiates control red block shall be functionally tested and cal-are given in Table 3.2-3. tbrated as inoicated in Table 4.2-1.

2. a. When a Limiting Control Rod <

Pattern extsts one of the Rod Block Monitors may be bypassed for maintenance and/or testing provided that this condition does not last longer than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in a 30 day period. If this condition lasts longer than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in a 30 day period the system shall be tripped.

The time spent while in a Limiting Control Red Pattern with one or more Red Block Monitors bypassed or inoperable and red withdrawal Dlocked does not count against the 24 nours in a 30 day period,

b. One enannel may be bypassed above 30* cower althout a time restriction provided that a Ltm1 ting Control Rod Pattern does not esist and the remaining Rod Block Monitor channel is operable.

c. Both Rod Block Monttor Channels are autenatically bypassed at less than 30%

rated thermal power or if the selected control red nas one or more adjacent fuel bundles Comprising the outer boundary of the reactor core. [

n 39a 3 . 2 <.t . 2 - 1 Amendment NO.

CUAO-CITIES OPR-30 Venturi tubes are provided in the main steamlines as a means of measuring steam flow and also Itmiting the loss of mass inventory from the vessel during a steamitne trear accident. In addit ten to monitoring steam flow, instrumentation is provided nich causes a trip of Group I tsolation valves. The primary function of the instrumentation is to detect a break in the main steamline, thus only Group 1 valves are closed. For the worst-case l accident, main steamitne break outside the drywell, this trip setting of 140% of r ated steam flow, in conjunction with the flew Itmtters and main steamline valve closure, limits the mass Inventory loss such that fuel is not uncovered, fuel terceratures remain less than 15000 F. and release cf radioactivity to the env ons is well below 10 CFR 100 guidelines (reference SAR Sections 14.2.3.9 and 14.2.3.10).

Temperature-monitor $ng instrumentation is provided in the main steamline tunnel to detect lears in this area. Trips are provided on this Instrumentation and nen excetoed cause closure of Group 1 1 solation valves. Its setting of 200 0 F is low enougn to detect leaks of the order of $ to 10 gpm; thus it 15 capaDie of covering the enttre spectrum of breaks. For large breats. It is a backup to nign-steam flow Instrumentation discussed above, and for small breaks with the resulting small release of radioactivity, gives isolatton efore the guideltnes of 10 CFR 100 are exceeded.

H1gh-radiation monit0rs in the main steamline tunnel have been proviced to detect gross fuel failure. This instrumentatten causes closure of Group 1 valves. the only val <es required to close for this accident. With the established setting :f 7 times normal backgr0und and main steamline isolation valve closure, fission product release is limited so that 10 CFR 100 guidelines are act exceeced f:r this accicent (reference SAR Section 12.2.1.7).

Pressure Instrumentat10n is proviced .nien trips nen main steam 1tne pressure drops belc. 325 asig. A trip of this instrumentation results in closure of Group 1 ' solation <alves. In the Refuel and Startup/ Hot Standey modes this trip funetten is Dypassed. This function is provided primartly to provide protection against a pressure regulator malfunction .hten would cause the c:ntrol an:/cr Oypass valve to open. With the trip set at $25 psig. Inventory loss 15 limited 50 that fuel 15 not uncovered and peak cladding temperatures are muen less than 15000 F; thus, there are no fissten products availacle for release other than those in the reactor water (reference SAR Section 11.2.3).

The RC1C and the HPC1 ntqn flow and temperature instrumentation are provided to detect a break in their respective piping. Tripping of this instrumentation results in actuatton of the RC1C or of HPC1 1solatien valves. Tripping logic for this function is the same as that for the matn steamline isolation <alves. thus all sensors are recutred to be c:erable or in a tripped conditi:n to meet single-failure criteria. The trip sett1ngs of 200*F and 300% of :esign f1cw and valve closure time are suen that :cre uncovery is prevente: and fission product release is within limits.

The Instrumentation nten initiates ECCS action is arranged in a one-out-of-t.o tauer t 1ce logic circuit. Unlike the reactor s: ram circuits, he.ever. t ere is one trio system associated with eacn function rather than the t.o trt: systems in the reacter protection system. The single-failure crite-ta are met Oy virtue of the fact that reduncant c:re c oling funct*:ns are :reviced. e.g. sprays and aut:matic D1c.cewn and hign-pressure :0:1 art in;ection. 4e s ectf1catt:n requires that if a trio system :ecomes r::e a:1e, tne system nten it activates is ce:lare tre er3 Die. For cia-0 1 e. if the tris system for : ore spray a ee:Omes inoperacle, core spraf a ts :e:larec incoera:1e and the cut-of-serv':e specifications of 5:e:*ft:ation 3.5 govern. This speciftcatten = reserves the effect1<eness of tre system .ith res:ect to the single-failure criteria even during :ertccs -rei maintenance or testing is cetng perforted.

i 1139H 3.2/4.2-6 Amendment NO.

[

e gua0-CITIES OFR-30

, .The,centrgl rnd block functions are provided to prevent excessive control rod withdrawal so that MCPR does not go celow the MCPR Fuel Cladding Integrity Safety Limit. The trip logic for this function is one out of n; e.g., any trip on one of the six APRN's, eight IRM's, four SRM's will result in a rod bicck. The minimum instrueent channel requirements assure sufficient instrumentation to assure that the $1ngle-fallure criteria are eg t . The minimum instrument channel requirements for the RBM may ce reouced by one for a short period of time ah11e in a Ltmiting Control Rod Pattern to allow for maintenance, testing, or calibration. This time period is only

, - 3% of the operating time in a month and does not significantly increase the risk of preventing an inadvertent control rod withdrawal. In addition, wn11e the unit is operating in a limiting control pattern with one or more Rod Block Monitors bypassed and control rod withdrawl blocked, this time does not count towards the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in a 30 day restriction. This time restriction is placed on the Rod Block monitor system to decrease the probability of a Rod Withdrawl Error wn11e in a limiting control red pattern. With control rod withdrawl clocked all rod withdrawl 15 prevented.

hence the RBM is not recuired to function.

1139H 3,2/4.2-64 4mendment 50.

. _ _ _ _ - -n

d~

, QUAD-CITIES OP3-30 The APRM roG block function is flow biased and prevents a significant reduction in MCPR. especially during operation at reduced flow. The APRM provides gross core protection. i.e., limits the gross withdrawal of control rods in the normal. withdrawal sequence.

In the refuel and startup/ hot standby modes, the APRM rod block function is set at 12% of rated power. This control rod block provides the same type of protection in the Refuel and $tartup/ Hot Standby modes as the APRM flow-blased rod block does in the Run mode, i.e.. prevents control rod withdrawal before a scram is reached.

The RBM rod block function provides local protection of the core. i.e., the prevention of transition bolling in a local region of the core for a single rod withdrawal error f ran a limiting control rod pattern. The trip point is flow biased. The worst-case single control rod withdrawal error is analyzed for each reload to assure that, with the specific trip settings, rod withdrawal 15 blocked before the MCPR reaches the fuel cladding integrity safety limit.

Below 30% power. the worst-case withdrawal of a single control rod without rod block action will not violate the fuel cladding integrity safety Itmit.

Thus the RBM rod block function is not required below this power level. If a control rod is selected that has one or more adjacent fuel bundles comprising the outer coundary of the reactor core, the neutron leakage is sufficiently high such that withdrawal of this rod will not violate the fuel cladding integrity Safety limit. Thus the RBM function is not recuired for control rods that nave one or more adjacent fuel bundles comprising the outer coundaries of the core.

The IRM block function provides local as well as gross core protection. The scaling arrangement is sucn that the trip setting is less than a factor of 10 above the indicated level. analysis of the worst-case accident results in rod block action before MCPR approacnes the MCPR fuel cladding Integrity safety Itmit.

A downs: ale indication on as AfRM is an indication the instrument has failed or is not sensitive enough. In either case the instrument will not rescend to changes in control rod motion, and the control rod motion is thus preventad. The downscale trips are set at 3/125 of full scale.

The $RM rod block wi',h 1 100 CPS and the detector not full inserted assures that the $RM's are not withdrawn from the core prior to commencing rod withdrawal for startup. The scram discharge volume high water level block provide annunciation for operator action. The alarm setpoint nas been selected to provide adequate time to allow determination of the cause of level increase and corrective action prior to automatic scram initiation.

For ef fective emergency core cooling for small pipe breaks the HPCI syst&9 must function since reactor pressure does not cecrease rapidly enough to allow either core scray or LPCI to operate in time. The automatic pressure relief function is provided as a backup to the NPCI in the event the HPCI does not ocerate. The arrangement of the tripping contacts is such as to provide this function nen necessary and minimi:e s:uricus coeration. The trip settings given in the specification are adequate to assure the acove criteria are met (reference SAR Section 6.2.5.3). The s:ecification preserves the ef fectiveness of the systen curing ;eriods of taintenance, testing or calibration and also 9inimi:es tre risk of inadvertent ::eration.

i.e. , only one instrument channel Out of service.

l Two radiation mcnitors are grovides :n the refuel 1Pg floor .nien 'nitiate 1 solation of the re3ctor builcing and c:eration of the stancey ;as treatment systems. The tric logic ts one Out of t o. Trip settings of 100 ma/nr for the ment ters on the ref;eling floor are cased ucer, initiat tng normal ventilation isolation and stancey gas treatment system o;eration Amendment NO.

1139H 3.2/4.2-7

e-

• *'. QUAD -C I TIE S OPR 30 TABLE 3.2-3

~

• NSTRUMENTATION THAT INITIATES 200 BLOCK Minimum Numeer

. of Operaele or Tripped Instruirent Channels per

-Trio Srstem C Pstrument fric level settina 2 APGM upscale (flow bias)( ) rop (2) 1(0.58wo + 50]

MFLPD 2 APtM Upscale (Refuel and 112/125 full scale Startup/ Hot Standby mode) 2 APRM downscale57I 11/125 full scale 1

Red 31oegntterupscale(flow 3145) 10.65We + 42(2) 1 R0d 31ock monttor downscale(7) 13/125 full scale 3 IRM downscale(3) (3) 11/125 full scale 3 IRM upscale (3) 1108/125 full scale 2(5)

$RM etector cosition ( g t in Startuo 12 feet below core centerline 3

IR"cetectoraStinStartuo cosition (3 12 feet below core centerline  !

2(5) (6) $as ;ose31e gio5 counts /see 2(5) $gg :ewnse33e W 1 2coyngsfsee 10 1 (per bank) Mtgn water level in scram i 25 gallons (per :ank) disenarge volume (SDV) 1 SDV hign .ater level scram NA trio eycassed

. '8 c t e s

1. For the Startuc/Wot Standby and Run positions of the reactor mode selector

' switen, there small te two operable or trioced trip systems for eacn function encept Pe SRM rod blocks. IRM upscale and IRN downscale need not i

be coerable in the tun position. APRM doanscale. APRM upscale (flow biased).

and RBM downscale *eed not be coeraole In the Startuo/ Hot Standey moce. The l EEM upscale and ::.nscale trips need not ce operable at less than 30; ratec thermal power or at any ocwer level if the selected control rod nas .ere er more adjacent fuel tundles comortsing the outer coundary of the reactor core. The RBM is aut:matica117 bypassed at less than 30 rated tnermal power or if tre se'ected c:ntrol rod has one or more adjacent fuel : uncles ccrorising the outer Cundary of the reactor core. For systems dith more than one caan9e1 :er trio system. If the first column cannot te met for ore of the two tric systems, this c:ndition may exist for up to 7 cays crovided tnat :uring that t we the :ceracle system ts functionally tested w edtatel y and daily thereafter: if tnis c:ndition lasts longer tnan 7 days tre system snail :e tricced. If tre first column cannot ce met for botn trio systems.

tne systems sna11 :e tripped.

1139H 1.2/4.2 14 Amendment NO.

'c QUAO-CI?!ES OPR-30

3.  ?$e control ro6 Brive nousing 3. The corrgetness of the control support system shall be in place rod withGrawal secuence input to during reactor power operation the RWH computer shall be vert-and when the reactor coolant fled after leading the sequence, system 1s pressurtzed above

. atmospneric pressure with fuel in the reactor vessel, unless Prior to the start of control rod withdrawal towards critical-all control rods are fully ity. tha capability of the rod inserted and Specificattnn worth minim 12er to properly ful-3.3.A.1 15 met, fill its function shall be vert-fied by the followino checks:

a. Control rod withe awa1 se-quences shall be ista,- a. The RWM computer on line 11shed so that ma.stmw1 diagnostic test shall be reactivity that etuld be successfully performed.

added by dropout o" a y in-crement of any one o' trol D. Proper annunctation of the blade would be suct tbat the selection error of one out-rod drop accident disa ;n of-sequence control rod Itmit of 230 cal /gm 11 not shall be verified.

exceeded.

b. Whenever the reactor in c. The rod block function of the $tartup/ Hot Standt; or the RWH Shall be verified by Run moce elcw 2*% rated withdrawing the first rod as thermal po.er, the rod nortn an out-of-secuence control mintm1zer snall :e rod no more ihan to the operable. A sec:nd co- block point.

erator or ualified tech-nical person may te used as a. Prior to control rod withdrawal a substitute for an Inocer- for startup or during refueling.

able rod north mintm12er verify that at least two source which fails after with- range channels have an CDserved drawal of at least 12 con- count rate of at least three trol rods to the 'ully counts per second.

withdrawn position. The rod worth minimizer may also be $. When a limiting control rod pat-bypassed for low ;ower tern ex tsts an instrument func-physics testing to tional test of the RBM shall be denonstrate the smutdown performed prior to withdrawal of margin require"ents of the designated rod (s) and daily Specification 3.3.A If a thereafter, nuclear engineer is present and verifies the steo- 6. The scram discharge volume vent by-steo rod movement s of the and drain valves shall be vert-test proredure. fled open et least once per 31 days. These valves may te

a. Control rods snall set te with- closed intermittently for test-dra.n for startup or refueling ing under administrative control unless at least two source range and at least once per 92 days, enannels have an ceserved count each valve snall be cycled rate equal to or greater than through at least one'comolete three c unts *er seq:nd and cycle of full travel. At least thess 5 M's are fully inserted, once each Refueling Outage, the scrim disenar;e volume vent and

5. Esceot as rovi ed :y drain valves will be

$pectricat!on 3.2.C.1 and demonstrated to:

3.2.C.2 suring :.aarata:n ith 11mittag c:nte:1 r: :atterns. a. Close within 30 sec nds af-as :etermiaed :p the *. clear ter receipt of a signal for engineer, ett9er: control rods to scram, and

4. octn RSM cnannels snall be C. Ocen amen the scram s1gnal ocersete. Is reset.

3 centrol rod <1tn:ra 41 shall be elected; er l

l Amendment NO.

i 1139H 3.3/a.3-3

e.

• QUAD-CITIES OPR-30

^

' b. the delayed neutron fraction chosen for the bounding reactivity curve

c. a beginning-of-life Doppler reactivity feedback

d. scram times slower than the Technical Specification rod scram insertion rate (Section 3.3.c.1)

e. the maximum possible rod drop velocity of 3.11 fps

f. the Jesign accident and w ram reactivity shape function, and

g. the moderatcr temperature at which criticality occurs In most cases the worth of insequence rf,Js or rod segments in conjunction with the actuai values of the other important accident analysis parameters described above, would most likely result in a peak fuel enthalpy substantially less than 280 cal /g design limit.

Should a control drop accident result in a peak fuel energy content of 280 cal /g. fewer than 660(7 x 7) fuel rods are conservatively estimated to perforato. This would result in an offsite dose well below the gu1Jeline value of 10 CFR 100. For 8 x 8 fuel, fewer than 850 rods are conservatively estimated to perforate, with nearly the

.,ame consequences as for the 7 x 7 fuel case because of the rod power differences.

The rod worth minimizer provides automatic supervision to assure that out of sequence control rods will not be withdrawn or inserted; i.e.. it limits operator deviations from planned withdrawal sequences (reference SAR Section 7.9). It serves as a backup to procedural control of control rod worth. In the event that the rod worth minimizer is out of service when required, a licensed operator or other qualified technical enployee can manually fulfill the control rod pattern conformance function of the rod worth minimizer. In this case, the normal procedural controls are backed up by independent procedural controls to assure conformance.

4. The source range monitor ($RN) system performs no automtlic Safety system function i.e.. it has no scram function. It does provide the operator with a visual indication of neutron level. This is needed for knowledgeable and efficient reactor startup at low neutron levels. The consequences of reactivity accidents are functions of the initial neutron flux. The requirement of at letst 3 counts per second assures that any transi beginsatorabovetheinitialvalueof10'gnt,shoulditoccur, of rated pcwer used in the the analyses of transients from cold conditions. One operable SRM channel would be adcquate to monitor the approach to criticality using homogeneous patterns of scattered control rod withdrawal. A minimum of two operable SRM's is provided as an added conservatism.

5. The Rod Block Monitor (RBM) is designed to automatically prevent fuel damage in the event of erronenus rod withdrawal from locations of high power density during high power operations. During reactor o;erations with cartain limiting control rod patterns, the worst-case withd/awal of a single centrol rod could result in one or more fuel rods with MCPR's less than the MCPR fuel cladding integrity safety limit. During a Limiting contcol rod pattern, testing of the RBM system will assare its operability prior to l

withdrawal of such control rods. To f acilitate testing while in a l

limiting control rod pattern ene RBM may be bypassed. for brief periods of tirr.e to perform maintenance and/or testing without decreasing the reliability of the system, providad the other RBM is operable. Two REM channels are provided. Tripping one operable i channel will block erroneous rod withdrawal soon enough to prevent violation of tne MCPR Safety limit. It is the responsibility of the nuclear engineer to identify these limiting control rod patterns and

! the designated r'.ds either when the patterns are initially established or as they develop due to the occurrence of inoperable control rods in other thaa limiting pattern.

6. The operability of the Scram Discharge Volume vent and drain valves assures the proper venting and draining of the Volume. 50 that water accumulation in the Volurre does not occur. Tr.ese specifications provide for the periodic verification that the valves are open, and for the testing of these valves under reactor scram conditions during each Refueling Outage.

1139H 3.3/4.3-9 l

ATTACHMENT _2

SUMMARY

OF CHANGES A total of twenty four (24) changes to the Quad Cities Station Units 1 and 2 Technical Specifications have been identified (12 per unit) and are listed below as follows:

1. Page 3.2/4.2-1, DPR-29 and 30 (a) Limiting Condition Operation (LCO) - Delete Item C.2 in entirety and replace with new items 2.a., 2.b., and 2.c..

This change is strictly a numbering change and is administratively in nature.

(b) LCO, Technical Specification 3.2.C. - Create new Item 3.2.C.2.a.,

which reads as follows: "When a Limiting Control Rod Pattern exists one or more...does not count against the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in a 30 day period."

This section was changed to clarify time restrictions on Rod Block Monitor bypassing. It now states that while in a limiting control pattern, one Rod Block Monitor may be bypassed for maini.enance/

testing for no longer than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in a 30 day period; unless in this condition rod withdrawal is blocked, then this time does not count against the 24 hrs. in a 30 day period.

(c) LCO, Technical Specification 3.2.C. - Create new Item 3.2.C.2.b.,

which reads as follows: "One channel may be bypassed above 30%

power. . .and the remaining ::od Block Monitor Channel is operable."

This section was changed so that if a limiting control rod pattern does not exist, then one Rod Block Monitor may be bypassed for any length of time, provided the other Rod Block Monitor is operable.

Rod Withdrawal nee 3 be blocked in this case.

(d) LCO, Technical Specification 3.2.C. - Create new Item 3.2.C.2.c.,

which reads as follows: "Both Rod Block Monitor Channels are automatically bypassed at less...of the reactor core."

This section would permit both Rod Block Monitors to be automatically bypassed below 30% rated thermal power and on edge control rods.

2. Page 3.2/4.2-6, DPR-29 and 30 (a) Delete last paragraph of Bases.

This paragraph to the Bases is being dropped and replaced with a new section. The new section encompasses the changes which result from the proposed clarifications to the Rod Block Monitor Technical Specifications.

.o,.

3. page 3.2/4.2-6a, DPR-29 and 30 (a) Create new pcragraph which reads as follows- "The Control Rod Block functions are provided to prevent...hence the RBM is not required to ,

function." )

This new section provides clarification that while in a limiting control rod pattern, one Rod Block Monitor may be bypassed for a short period of time to perform maintenance and/or testing. This provision is currently in the Bases, however, would otherwise have been deleted based on the change described in Item 2(a). In addition, while the unit is operating in a limiting control pattern with one or more Rod Block Monitors bypassed and control rod with-drawal block, this time does not count towards the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in a 30 day restriction.

4. Page 3.2/4.2-7, DPR-29 and 30 (a) Bases - fourth paragraph, third line. Following the sentence which ends with the words "power level.", insert the following: "If a control rod is selected that has one or more adjacent fuel bundles...the outer boundaries of the core".

This addition to the fourth paragraph clarifies that both Rod Block Monitors are automatically bypassed below 30% power and when a control rod, with one or more fuel bundles residing on the reactors perifery, is selected.

5. Page 3.2/4.2-14, DPR-29 and 30 (a) Note 1, fourth line - Following the words "RBM upscale", insert the words "and downscale trips", so that the sentence now reads "The RBM upscale and downscale trips need be operable at less than 30% rated thermal power,"

(b) Note 1 (continued) - Following the words "at less than 30% rated i thermal power," insert the words "or at any power level if the selected control rod... comprising the outer boundary of the reactor core" so that the sentence now reads, "The RBM upscale and downscale j trips...or at any power level...of the reactor core."

i j (c) Note 1 (continued) - Dc <te the sentence "One channel may be bypassed

! above 30% rated ther'< ' rower provided that a limiting control rod l pattern does not ex!, .

These changes (Sa, 5b, and Sc), provide clarification as to when the l

l upscale and downscale trips of the Rod Block Monitoc are not required l and when they are automatically bypassed.

1 l

l l, _ _

6. Page 3.3/4.3-3. DPR-29 and 30 (a) LCO, Technical Specification 3.3.B.5 - Insert the words "Except as provided by Specification 3.2.C.1 and 3.2.C.2" so the sentence now reads "except as provided by the by Specification 3.2.C.1 and 3.2.C.2 during operation with a limiting control rod patterns..."

This addition clarifles that while in a limiting control rod pattern, one Rod Block Monitor may be bypassed for maintenance and/or testing. In addition, it also clarifies that an edge rod may have both Rod Block Monitors bypassed while in a Limiting Control Rod pattern. ,

7. Page 3.3/4.3-9. DPR-29 and 30 (a) Bases, Item 5, third line - Delete the section beginning with the words, "Two channels are provided..." thro.gh the sentence containing the words "identify these limiting patterna". Replace with section that begins with words "During reactor operations with certain limiting control rod patterns...." through the sentence that contains the words "identify these limiting control rod patterns".

This addition clarifies the Rod Block Monitor operability requirement during a limiting control rod pattern.

1 1

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l 4637K l

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

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

4.4 L , ATTACHMENT 3

.)

l l

BASIS FOR SIGNIFICANT HAZARDS CONSIDERATION CLARIPICATION OF ROD BLOCK MONITOR (RPM)

OPERABILITY AND BYPASS TIME REOUIREMENTS The proposed Technical Specification amendments to sections 3.2.C, Table 3.2-3, 3.3.B.5 and the Basis for these sections, are being submitted to clarify interpretations of these sections. This Technical Specification amendment will clarify bypass time limitations on the Rod Block Monitors and also Rod Block Monitor operability.

The proposed Technical Specification amendment allows 1 Rod Block Monitor to be bypassed without any time limitations provided the other Rod Block Monitor is operable and a Limiting Control Rod pattern does not exist.

One Rod Block Monitor is capable of preventing the worst case unrestricted rod withdrawl from violating the MCPR Safety Limit. By definition a Limiting Control Rod pattern is a condition in which the worst case, unrestricted withdrawl of a control rod could violate the MCPR Safety Limit. However, due to changing conditions with core flow, xenon, or control rod movement it is difficult to determine when the unit enters a limiting control rod pattern.

Therefore providing for one Rod Block Monitor to be operable while not in a limiting control rod pattern will prevent the worst case unrestricted rod withdrawal if a limiting control rod pattern is entered.

While the unit is operating in a Limiting Control rod pattern both Rod Block monitors must be operable, except for maintenance and/or testing in which case one may be bypassed for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in 30 days. This short period of time will allow proper testing of the Rod Block Monitors and will not significantly increase the risk of the worst case unrestricted rod withdrawal to violate the MCPR Safety Limit.

In addition, while the unit is operating in a limiting control pattern with one or more Rod Block Monitors bypassed and control rod withdrawal block, this time does not count towards the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in a 30 day restriction. This time restriction is placed on the Rod Block monitor system i

to decrease the probability of a Rod Withdrawal Error while in a limiting

! control rod pattern. With control rod withdrawal blocked all rod withdrawal l is prevented, hence the RBM is not required to function. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> time

! should only apply when the RBM is required to provide a Rod Block Function.

1 l

1

.8, s

. When reactor power is less than 30% rated core thermal power, or an edge rod is selected, at any power, both rod block monitors are automatically bypassed. General Electric's Equipment Manual APED-5706, November 1968, documents that below this power level the bundle powers are low enough that withdrawal of the strongest rod will not violate the MCPR Safety Limit.

Analysis also showed significant neutron leakage on the edge fuel assemblies of the core such that withdrawal of any edge control rod will not violate the MCPR Safety Limit.-

These changes have been reviewed by Commonwealth Edison and we believe they do not present a Significant Hazards Consideration. The basis for our determination is documented as follows:

BASIS POR NO SIGNIFICANT HAZARD DETERMINATION commonwealth Edison Ccapany has evaluated thic proposed amendment and determined that it involves no significant hazards cons'derations. In accordance with the criteria of 10 CFR 50.92(c), a propored 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 proposed Technical Specification changes require at least one Rod Block Monitor to be operable whenever the worst case unrestricted rod withdrawal error could violate the MCPR Safety Limit. One Rod Block monitor is sufficient to prevent the MCPR Safety Limit from being violated during the worst case, unrestricted rod withdrawl.

Therefore, this does not involve a significant increase in the probability or consequence of an accident previously evaluated.

2) Create the possibility of a new or different kind of accident from any accident previously evaluated; because; (a) A review of the proposed Tecnnical Specification changes does not reveal a new or different kind of accident from any previously evaluated. This proposed amendment does not change the times the Rod Block monitor is needed and therefore does not create the possibility of a new or different kind of accident than previously was evaluated, l

7. *

3) Involve a significant reduction in the margin of safety, because; (a) The proposed Technical Specification change requires at least one Rod Block Monitor to be operable when reactor power is sufficient so that the worst case unrestricted rod withdrawl could violate the MCPR Safety Limit. The only time that the worst case unrestricted rod withdrawl could violate the MCPR Safety Limit is when the unit is operating in a Limiting Control rod pattern. However, due to changing conditions with core flow, Xenon, or control rod movement it is difficult to determine when the unit enters a limiting control rod pattern.. Therefore, providing one Rod Block Monitor to be operable while not in a limiting control rod pattern will prevent the worst case unrestricted rod withdrawl if a limiting control rod pattern is entered. In addition when a limiting control rod pattern exists both Rod Block Monitors must be operable or rod withdrawl be blocked, except for maintenance and/or testing for brief periods of time.

This ensures with a high degree of certainty that the worst case rod withdrawl error will not violate the MCPR Safety Limit. Hence, the changes do not result in a decrease in the margin of safety.

Therefore since the proposed license amendment satisfies the criteria specified in 10 CFR 50.92, Commonwealth Edison has determined that a no significant hazards consideration exists for this license amendment. We request its approval in accordance with the provisions of 10 CFR 50.91(a)(4).

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